Dispersion-strengthened Tungsten Research Articles

Effect of dispersed particles on microstructure evolution and mechanical properties of 93W–4.6Ni–2.1Fe–0.3Co alloys

93W–4.6Ni–2.1Fe–0.3Co alloys reinforced by ZrC particles were prepared by the two-step sintering method in this article. The conventional sintering process conducted at a temperature of 1485 °C resulted in the formation of microvoids between the larger dispersed particles and the matrix, thereby leading to a degradation in the tensile property (959 MPa/11.5%). Through controlling the sintering temperature at 1440 °C, a dispersion-strengthened tungsten heavy alloys (WHAs) with excellent comprehensive tensile properties (974 MPa/27.0%) was obtained, mainly attributed to the synergistic effect of homogeneous dispersion strengthening and grain boundary purification. Under high compressive strain rates, dispersion-strengthened WHA demonstrates a superior yield strength compared to WHA, suggesting that dispersion-strengthened WHAs have potential applications in the armour-piercing field.

Restoration kinetics and microstructural evolution of moderately warm-rolled yttria dispersion-strengthened tungsten plate during annealing at high temperatures

The thermal stability of an yttria particle dispersion strengthening tungsten (W-2vol%Y2O3) plate with 67 % rolling thickness reduction (WY67) fabricated by powder metallurgy and warm rolling is investigated through isochronous annealing and isothermal annealing. Hardness testing of annealed specimens allows tracking the degradation of the mechanical properties and electron backscatter diffraction (EBSD) was used to characterize the corresponding microstructure. Isochronous annealing for 1 hour between 1150 °C and 1450 °C indicates that the recrystallization temperature of WY67 is about 1400 °C. Herein, the recovery kinetics, recrystallization kinetics, and the evolution of microstructure and texture of WY67 during isothermal annealing at temperatures ranging from 1300 ℃ to 1450 ℃ were investigated in detail. The effect of both the heterogeneity of the deformed microstructure and the yttria particles on the recovery and recrystallization was further studied. The results show that the stored energy of the deformed grains of yttria dispersion strengthened tungsten is texture-dependent and the grain with γ-fiber (normal direction // <111>) texture has the largest stored energy of 546 KJ/m3. The grains with higher stored energy are conducive to recrystallized nucleus/grain growth rather than nucleation during subsequent annealing. The recrystallized texture inherits the deformed texture and is mainly composed of γ-fiber texture. There are more pre-existing potential recrystallization nuclei in the elongated grains with larger deformation in the as-rolled microstructure. Micron-sized yttria particles can stimulate the formation of recrystallization nuclei through particle-stimulated nucleation mechanism, nevertheless, this is not the dominant nucleation mode for WY67 during recrystallization in this study. Fine (submicron) yttria particles hinder the recovery and recrystallization processes through the Smith–Zener pinning effect.

Recrystallization kinetics of larger yttria particle dispersion tungsten alloy warm rolled to 50% thickness reduction

Tungsten-based materials are considered one of the candidate materials for the key components (such as the first wall and diverter) of future thermonuclear fusion reactors because of their excellent properties. At high operation temperatures, restoration processes such as recovery, recrystallization and grain growth will unavoidably embrittle ductile tungsten achieved by rolling. Dispersed second-phase particles will affect the recrystallization behavior. The thermal stability of a yttria dispersion-strengthened tungsten plate warm-rolled to 50% thickness reduction is investigated. Isochronous annealing is used to predict its recrystallization temperature range and isothermal annealing is used to study its recrystallization kinetics. The tungsten alloy in the present study has a faster recrystallization kinetics and a smaller activation energy compared to that of W-2 vol% Y2O3 alloy with small Y2O3 particles, indicating that the larger Y2O3 particles have little effect on impeding grain boundaries during recrystallization.

Proof-of-Concept Modeling for the Use of Neutron Reflectometry for the In Situ Diagnosis of Deuterium Accumulation in Tungsten and Dispersion-Strengthened Tungsten Alloys

The diagnosis of plasma-facing components in a fusion environment is challenging due to the limited number of measurement techniques that have been developed for in situ surface analysis. In this work, we assess the feasibility of using neutron reflectometry (NR) for the in situ diagnosis of deuterium accumulation in tungsten and dispersion-strengthened tungsten alloys. TRIM is used to simulate deuterium implantation at different energies to approximate the deuterium depth profiles in these materials in order to calculate the expected measurements from NR for various fluences. Our results suggest that NR should be an effective technique for testing hypotheses about the surface composition of materials under fusion-relevant fluences of deuterium irradiation.

Effects of transition metal carbide dispersoids on helium bubble formation in dispersion-strengthened tungsten

The formation of helium bubbles and subsequent property degradation poses a significant challenge to tungsten as a plasma-facing material in future long-pulse plasma-burning fusion reactors. In this study, we investigated helium bubble formation in dispersion-strengthened tungsten doped with transition metal carbides, including TaC, ZrC, and TiC. Of the three dispersoids, TaC exhibited the highest resistance to helium bubble formation, possibly due to the low vacancy mobility in the Group VB metal carbide and oxide phases. Under identical irradiation conditions, large helium bubbles formed at grain boundaries in tungsten, while no bubbles were observed at the interfaces between the carbide dispersoid and tungsten matrix. Moreover, our results showed the interfaces could suppress helium bubble formation in the nearby tungsten matrix, suggesting that the interfaces are more effective in trapping helium as tiny clusters. Our research provided new insights into optimizing the microstructure of dispersion-strengthened tungsten alloys to enhance their performance.

The W-Zn-Co-Y2O3 alloys synthesized by a secondary ball milling method and their effects on adhesion performance of single lap joints of aluminum composites

W-Zn-Co-Y2O3 tungsten heavy alloys were produced using a two-step mechanical milling method. In the first step, Zn-Co raw materials were milled for 24 h. Then, in the second stage, the Zn-Co compounds obtained were mechanically milled with tungsten (W) and yttrium(III) oxide (Y2O3). This study compared the new W-Zn-Co-Y2O3 alloy, obtained through two-stage mechanical alloying, with the same alloy produced using classical mechanical milling. The aim was to investigate and compare their structural, morphological, and mechanical properties. Yttrium oxide was used to promote the formation of in-situ oxide dispersoids during mechanical alloying. Oxide dispersion-strengthened tungsten heavy alloys are known for their exceptional mechanical properties, making them suitable for various high-temperature applications. The structures were analyzed using XRD, crystallite size, TEM, SEM, and EDX techniques. The microstrain of the final alloys was calculated as 15.79 × 10−3 and 13.12 × 10−3 for the classically obtained alloy and secondary milled alloys after 24 h of milling, respectively. Additionally, the reinforcing effects of the produced alloy on single-lap joints of aluminum composites were investigated. The results demonstrated that the tungsten alloy produced using the secondary ball method exhibited better mechanical performance.

Examination of Early-Stage Helium Retention and Release in Dispersion-Strengthened Tungsten Alloys

Tungsten is the material of choice for plasma-facing components in the divertor region of future nuclear fusion reactors. Exposure to low-energy helium ion irradiation results in microstructural changes as helium is trapped at defects in the tungsten matrix. High-temperature exposure results in the formation of helium bubbles in the subsurface. Dispersion-strengthened tungsten materials are tungsten-based materials with added transition metal carbides to alter the impurity distribution and grain structure. In this work, the thermal release of helium from dispersion-strengthened tungsten is investigated. After irradiation at 1073 K to a 1024 m−2 fluence, thermal desorption spectroscopy was performed to elucidate the helium trapping and desorption behavior. Post-desorption microscopy was performed to correlate the microstructural changes with helium release spectra. The amount of desorbed helium was highest in the 1.1 and 5 wt% alloys, and significantly lower in the 10 wt% alloys. Helium bubbles were observed in the pure tungsten and 1.1 wt% alloys within the tungsten grains. Correlating the composition with helium release spectra revealed the importance of tailoring grain size and oxide vacancy concentrations by varying the dispersoid content on the helium retention and release behavior. These first results of helium desorption from dispersion-strengthened tungsten indicate compositionally dependent retention and reveal the need to examine helium retention in advanced tungsten alloys under reactor-relevant exposure.

Inhomogeneous deformation substructure and its effect on the recrystallization behavior of yttria dispersion-strengthened tungsten plate

Tungsten materials have been chosen for use in critical parts (i.e., as the armor of plasma-facing material) in fusion reactors. Deformed tungsten will undergo recrystallization at a high-temperature service environment, and recrystallized tungsten exhibits brittleness, which reduces mechanical properties, thermal shock resistance and service lifetime. The deformation heterogeneity may result in an inhomogeneous microstructure, affecting its recrystallization behavior. The main aim of the present work is to qualitatively and quantitatively characterize the heterogeneity in the through-thickness deformed microstructure of an yttria dispersion-strengthened tungsten plate warm-rolled to 50 % thickness reduction ratio and study the effect of heterogeneous microstructure on recrystallization behavior. For this, topological maps of the surface, quarter and center region are measured on the deformed state and annealed state utilizing electron backscatter diffraction (EBSD). Distributions of grain size, geometrically necessary dislocations (GND) energy, potential nuclei density and texture have been analyzed in detail. The results revealed that the deformed microstructure of tungsten plate along full-thickness has no apparent heterogeneity gradient of grain size and texture, but the center layer has a larger subgrain size and more potential recrystallized nuclei compared to the surface layer. The larger subgrain is more conducive to static recrystallization nucleation, while the existing potential recrystallized nuclei can grow rapidly without incubation time during subsequent high-temperature annealing, thereby recrystallization preferentially undergoes in the center layer of the tungsten plate.

Microstructure and mechanical properties of in-situ formed ZrC nanoparticles dispersion-strengthened tungsten alloy

W-Zr-C alloy with nanoscale ZrC particles dispersion was fabricated via powder metallurgy method using W, ZrH2 and nanoscale C powders as starting materials. The average size of in-situ formed particles is 55 nm. The smaller particles in the grain interior are dominantly ZrC particles. Zr decomposed from ZrH2 can also react with impurity oxygen to form ZrO2 particles and reduce the detrimental effects of oxygen on grain boundaries. The as-swaged W-Zr-C alloy is ductile at 200 °C, and the ultimate tensile strength and total elongation (TE) at 300 °C are 643.5 MPa and 23.5%, respectively. After annealing at 1400 °C, the UTS at 300 °C of W-Zr-C alloy is still as high as 611.4 MPa and the TE is 33.2%. The recrystallization start temperature of the as-swaged W-Zr-C alloy is between 1400 and 1500 °C, which is 200 °C higher than that of pure W. The in-situ formation of nanoscale second-phase particles via the dissolution-precipitation mechanism provides a feasible strategy for improving the low-temperature toughness and high-temperature stability of tungsten alloys.

Nanostructured TiC dispersion-strengthened tungsten composite with remarkably improved He ion irradiation resistance

Improving the irradiation resistance of W alloys is vital to their application of plasma facing materials in future fusion reactors. In this work, quasi-spherical tungsten nanoparticles synthesized by thermal plasma were mixed with nanosized TiC to sinter nanostructured TiC dispersion-strengthened tungsten alloy. As-obtained W-TiC composite exhibited high relative density of 99.14% with suppressed grain size of about 1.12 μm, together with well-dispersed TiC with size 50–100 nm into W grains. More importantly, semi-coherent W/TiC interface with diffusion layer of about 25 nm was formed, which highlights the tight connection between W and TiC. Furthermore, He ions irradiation results indicated that W-TiC composite exhibited excellent irradiation resistance with a damage depth of about 9.53 μm, which is much lower than 18.26 μm of pure W. Mechanism investigation indicated that the remarkably improved He ion irradiation resistance was attributed to the efficient capture and annihilation of irradiation-induced defects by grain boundaries and phase interfaces.

Improving the mechanical properties and thermal shock resistance of W-Y2O3 composites by two-step high-energy-rate forging

Yttria dispersion-strengthened tungsten composites with enhanced thermal and mechanical properties were fabricated by powder metallurgical methods in this work. The processing route involves the doping of yttria particles in the tungsten powders by liquid-liquid doping process, followed by hot pressing and two-step high-energy-rate forging (HERF). The microstructure, thermal conductivity, mechanical properties and thermal shock resistance of the hot-pressed W-Y2O3 composites and the HERF processed W-Y2O3 composites were comparatively investigated. The bending strengths of the HERF processed W-Y2O3 composites increased significantly compared with that of the hot-pressed W-Y2O3 billets. The HERF W-Y2O3 material presented a ductile behavior even at room temperature with a maximum flexural strain of 4.4% and a yield stress of 2324.5 MPa. At 100°C, the material also exhibits evident plasticity with a flexural strain of 5.9% and the yield stress is up to 1931.9 MPa. The thermal conductivity of the W-Y2O3 composite is 170 W/mK, which is close to the deformed pure W. The thermal shock response of the HERF processed W-Y2O3 is also evaluated by applying edge-localized mode like high heat loads (100 pulses) with an energy density of 0.16–1 GWm−2 at various temperatures in an electron beam facility. The thermal shock results indicate that the cracking threshold of the HERF processed W-Y2O3 at RT is between 0.55 GWm−2 and 0.66 GWm−2. The cracking threshold of the HERF W-Y2O3 at 100°C and above is higher than 1 GWm−2.

Transient thermal shock and helium ion irradiation damage behaviors of ODS-W/CuCrZr joints

The transient thermal shock and helium ion irradiation damage behaviors of oxide dispersion-strengthened tungsten (ODS-W)/CuCrZr alloy joints prepared by direct diffusion bonding and by improved diffusion bonding via surface nano-activation and Cu electroplating were studied. The surface morphologies and interface structures of the ODS-W/CuCrZr joints after thermal shock and He+ irradiation were investigated by atomic force microscopy, scanning electron microscopy, focused ion beam/scanning electron microscopy, and high-resolution transmission electron microscopy. The joint of ODS-W and CuCrZr prepared by the improved diffusion bonding method exhibited superior resistance to crack formation. No obvious plastic deformation was found on the ODS-W surfaces of the ODS-W/CuCrZr joints obtained by either bonding method after transient thermal shock (power density = 0.6 GW/m2). Interfacial debonding was observed to occur at the interface of W and Cu in the direct diffusion-bonded ODS-W/CuCrZr joint where Y2O3 particles exist, but did not occur in the joint prepared using the improved method. The nanostructures formed at the interface of the joint via the improved diffusion bonding method therefore elevated the radiation resistance of the joint. These results should be of relevance for the development of plasma-facing components in future nuclear fusion reactors.

Evolution of microstructure and texture in a warm-rolled yttria dispersion-strengthened tungsten plate during annealing in the temperature range between 1200 °C and 1350 °C

The thermal stability of an yttria dispersion-strengthened tungsten plate warm-rolled to 50% thickness reduction is investigated. Isothermal annealing experiments were conducted at four temperatures and the microstructure and texture evolution were analyzed with Transmission Electron Microscopy (TEM) and Electron Backscatter Diffraction (EBSD) to understand the restoration mechanisms in the W-2 vol% Y2O3 alloy. At 1200 °C, dominantly recovery occurred in the warm-rolled plate up to the longest annealing time of 84 h, whereas at temperatures of 1250 °C and above recrystallization was the dominant restoration mechanism. Microstructure and texture evolution during recrystallization were quantified in terms of grain size, aspect ratio and volume fractions of fiber texture components. The analysis reveals that the Y2O3 particles play a key role in the recrystallization mechanism of the plate affecting both, nucleation and growth behavior. Their dominating effect originates from pinning of high angle boundaries causing no-convex grain shapes as well as slower recrystallization kinetics than pure tungsten, thereby improving the thermal stability of the oxide dispersion-strengthened material.

A study on the effect of solution acidity on the microstructure, mechanical, and wear properties of tungsten alloys reinforced by yttria-stabilised zirconia particles

Solution acidity greatly affects the fabrication of high-performance oxide dispersion-strengthened tungsten (ODS-W) alloys by liquid-liquid doping. In the present study, yttria-stabilised zirconia (Zr(Y)O2)-strengthened tungsten alloys (W-Zr(Y)O2) were developed via azeotropic distillation. The effect of solution pH on the precursor morphology and microstructure and the mechanical properties of the corresponding alloys was investigated. When the solution pH increased from 2 to 8, a larger number of bonding Zr(Y)O2 particles could be observed in the alloys. The W-Zr(Y)O2 alloy fabricated with the pH 2 precursor powder exhibited a better microstructure and mechanical properties (relative density, micro-hardness, yield, and ultimate strength) than the alloys prepared with pH 5 and 8 precursor powders. The ultimate compressive strength and failure strain of the pH 2 alloy were 1009 MPa and 0.22, respectively. In addition, this alloy contained finer particles than the pH 5 and 8 alloys. The grain sizes of W-Zr(Y)O2 alloys fabricated with pH 2, 5, and 8 precursor powders were in the range of 2–6 μm, which implies that they were much smaller than the particles in pure tungsten. Due to a complicated abrasion mechanism, the wear resistance of the W-Zr(Y)O2 alloys initially increased and then decreased with an increase in the Zr(Y)O2 content. Finally, we could confirm that the alloy fabricated with the pH 2 precursor powder (with 3.0 wt.% Zr(Y)O2) exhibited the most optimal chemical composition, microstructure, and compressive properties. In addition, the wear resistance of the W-Zr(Y)O2 alloy fabricated with the pH 2 precursor powder was characterised. Finally, we expect that the results outlined in this study will act as basic guidelines for the fabrication of ODS-W alloys by liquid-liquid doping.

Nanostructured TiC Dispersion-Strengthened Tungsten Alloys with Remarkably Improved Irradiation Resistance

Nanostructured TiC Dispersion-Strengthened Tungsten Alloys with Remarkably Improved Irradiation Resistance

Mechanical properties and thermal shock resistance of tungsten alloys strengthened by laser fragmentation-processed zirconium carbide nanoparticles

Zirconium carbide (ZrC) nanoparticles with an average size of 5.6 nm were synthesized through laser fragmentation (LF) from as-received 20–60 nm ZrC particles, and LF-ZrC nanoparticle dispersion-strengthened tungsten (LF-WZC) samples were fabricated by spark plasma sintering method. The average grain size of LF-WZC is 1.91 μm and most ZrC particles in LF-WZC are smaller than 10 nm. LF-WZC exhibits finer grain size, higher yield strength and hardness but lower ductility as compared with W-ZrC samples using as-received ZrC (WZC). The results showed that finer ZrC nanoparticles dispersed in tungsten can enhance the strength by hindering the motion of dislocations, but they may also introduce stress concentration and thus reduce the ductility. The thermal shock resistance of the WZC and LF-WZC samples was investigated using an electron beam device. The LF-WZC sample also exhibits a higher cracking threshold (0.33–0.44 GW·m−2) than WZC (0.22–0.33 GW·m−2) at room temperature. The enhanced thermal shock resistance of LF-WZC could be attributed to its high yield strength.

High Flux Helium Irradiation of Dispersion-Strengthened Tungsten Alloys and Effects of Heavy Metal Impurity Layer Deposition

Tungsten has been chosen as the plasma-facing material (PFM) for the divertor region in ITER and also a candidate PFM for future plasma-burning nuclear fusion reactors. During fusion device operation, PFMs will be exposed to low-energy He irradiation at high temperatures, resulting in sub-surface bubbles and surface morphology changes such as pores and fuzz. Carbide dispersion-strengthened W materials may enhance the ductility of W, but their behavior under high flux He irradiation remains unclear. In this work, the response of dispersion-strengthened tungsten materials to high flux, low energy He irradiation at high temperature is examined. Tungsten alloyed with 1, 5, or 10 wt. % tantalum carbide or titanium carbide exposed to these conditions result in surface pores, coral-like feature growth and sub-surface helium bubbles. Reactor-relevant helium irradiation (5x1026 m-2 fluence) combined with high powered laser pulses to simulate off-normal reactor events does not significantly alter the surface morphology, as the surface nanostructures appear stable and cracks are only observed on a localized region of one sample. However, specimens show the development of an impurity layer on the surface, likely impurity deposition from the sample holder during irradiation, resulting in a mixed material layer on the surface. Helium bubbles exist in this impurity layer, and obscure conclusions about helium interactions with the carbide dispersoids. Nonetheless, it is clear that the dispersoid microstructure limits He bubble formation and subsequent surface nanostructuring, attributed to the dispersoid composition.

Research on preparation process for the in situ nanosized Zr(Y)O2 particles dispersion-strengthened tungsten alloy through synthesizing doped hexagonal (NH4)0.33·WO3

In this article, the in situ nanosized Zr(Y)O2 particles uniformly distributed in the tungsten alloy was successfully prepared through synthesizing doped α-HATB powder and spark plasma sintering process. The processing route involves a molecular-level liquid–liquid doping technique that causes a large number of nanosized particles within tungsten grains. The synthesis mechanism of α-HATB and phase evolution were detailedly investigated. Over 75% of W–Zr(Y)O2 powders particles are less than 3 μm. The determined Sw values (5.7), coefficient of uniformity Cu (2.3) of the W-0.5% Zr(Y)O2 doped tungsten are 5.7 and 2.3, respectively, both indicating the narrower size distribution. The average size of Zr(Y)O2 particles in prepared W alloy are about 250 nm under SEM observation. Through milling of these powders, the particles in W-0.5%Zr(Y)O2 alloy and 92.25W-4.9Ni-2.1Fe-0.75ZrO2 can further decrease to less than 100 nm in size, which are 8–10 times smaller than those in the state-of-the-art review.

GD-OES study of the influence of second phase particles on the deuterium depth distribution in dispersion-strengthened tungsten

Dispersion-strengthened tungsten materials represent a new class of W-based materials to be investigated for use as plasma-facing component in nuclear fusion reactors. However, the retention and permeation characteristics of these materials under low energy deuterium (D) irradiation need to be elucidated before the efficacy of these materials can be judged. Due to possible deep penetration of D in W, depth profile techniques such as glow discharge optical emission spectroscopy (GD-OES) are needed to probe D concentrations many microns beneath the material surface. In this study, the D retention behavior of W materials containing 1–10 wt% TaC, TiC, or ZrC are investigated with GD-OES. After exposure to a 5 × 1018cm−2 D ion fluence at 100 °C, D was observed beyond the D implantation depth the surface in many specimens, and the D depth profile was found to depend on the type and concentration of the added second phase. Combined with in-situ X-ray Photoelectron Spectroscopy (XPS) studies, the effects of impurity oxygen atoms on the D retention is considered, as an increasing oxygen content correlates with decreased D retention. The influence of grain size, second phase particles, and oxygen content on the retained D depth and concentration in these complex W-based materials is discussed.

The thermal stability of dispersion-strengthened tungsten as plasma-facing materials: a short review

One key challenge for the development of fusion energy is plasma-facing materials. Tungsten-based materials are promising candidates for plasma-facing components (PFCs) in the magnetic confinement nuclear fusion reactors because of their high melt temperature, high-thermal conductivity, high-thermal load resistance, low tritium retention, and low sputtering yield. In fusion reactors, PFCs are exposed to high-thermal flux, because there are some transient events such as plasma disruptions, edge-localized modes, and vertical displacement events (VDEs). Especially, in VDEs, a heat flux of 10–100 MW m−2 with duration of milliseconds-to-several seconds can induce recrystallization and then change the microstructure of tungsten-based plasma-facing materials, leading to instability of microstructures. Then, a significant degradation of material properties is caused such as a reduction of mechanical strength and fracture toughness, a rise in the ductile-to-brittle-transition temperature well, and decrease of irradiation/high-thermal load resistance. Therefore, many efforts were devoted to improve the thermal stability of tungsten-based materials as high as possible, such as oxide dispersion strengthening, carbide dispersion strengthening, and K bubbles dispersion strengthening. Here, the thermal stabilities of various dispersion-strengthened tungsten materials are reviewed by evaluating their recrystallization temperature and the corresponding hardness evolutions. In addition, the possible development trends are proposed.