Promising Plasma-facing Material Research Articles

Investigation of W-SiC compositionally graded films as a divertor material

W-SiC composite material is a promising plasma-facing material candidate alternative to pure W due to the low neutron activation, low impurity radiation, and low tritium diffusivity of SiC while leveraging the high erosion resistance of the W armor. Additionally, W and SiC have high thermomechanical compatibility given their similar thermal expansion rates. The present study addresses the synthesis and performance of compositionally graded W-SiC films fabricated by pulsed-DC magnetron sputtering. Compositional gradients were characterized using transmission electron microscopy (TEM) and energy-dispersive X-ray spectroscopy (EDS), and crystallographic information was obtained using electron diffraction and X-ray diffraction (XRD). Samples were exposed to L-mode deuterium plasma discharges in the DIII-D tokamak using the Divertor Material Evaluation System (DiMES). Post-mortem characterizations were performed using scanning electron microscopy (SEM) and XRD. Electron diffraction and XRD showed that the compositionally graded W-SiC films were composed of polycrystalline W and amorphous SiC with amorphous W+SiC interlayers. No macroscopic delamination or microstructural changes were observed under mild exposure conditions. This study serves as a preliminary examination of W-SiC compositionally graded composites as a potential candidate divertor material in future tokamak devices.

Enhanced thermal-mechanical properties of rolled tungsten bulk material reinforced by in situ nanosized Y–Zr–O particles

Tungsten is the most promising plasma facing material for fusion reactors. Rolled W–Y2(Zr)O3 bulk material has been successfully produced in this study for future fusion engineering applications. The introduction of Zr is conducive to the refinement of the second phase particles. Nano-sized Y–Zr–O particles are observed in the powder and bulk samples. Related results show that the Y–Zr–O particle dispersion distribution improves the heat load resistance of W–Y2(Zr)O3 composite material. For four-point bend experiments in the same sampling direction, the DBTT of W–Y2(Zr)O3 composite materials is lower compared to the pure tungsten. For the same material, the DBTT of the material was selected for testing along the RD direction is lower compared to the material was selected for testing along the TD direction. Findings of this study provide suggestions for the subsequent industrial preparation of nanoscale particle-doped tungsten materials.

Morphological damage to tungsten eroded by abrasive water jet

Abrasive water jet is a promising method for processing rolled tungsten plates, which are considered the most promising plasma-facing material in fusion devices. To investigate the erosion of tungsten due to the abrasive water jet, we conducted erosion tests and numerical simulations. We examined the morphological characteristics of the tungsten erosion damage. An erosion model was developed by coupling smoothed-particle hydrodynamics with the finite-element method to simulate tungsten erosion. The mechanism behind the morphological damage characteristics of tungsten was analyzed. Our findings reveal that as erosion depth increases, the kinetic energy of the abrasive particles decreases while the erosion angle increases. The erosion damage to tungsten transitions from brittle fracture induced by microcutting and microplowing to impact deformation. As a result, the erosion section's morphology deteriorates, exhibiting strips from the cutting process, increasing both the trailing angle and surface roughness. The morphological damage characteristics of tungsten are intricately linked to the abrasive particles' motion. Therefore, by altering the motion characteristics of these particles, the erosion section's quality can be enhanced. This research offers valuable insights for enhancing the quality of tungsten sections processed by abrasive water jets in fusion devices.

The Influence of Coherent Oxide Interfaces on the Behaviors of Helium (He) Ion Irradiated ODS W.

Tungsten (W), as a promising plasma-facing material for fusion nuclear reactors, exhibits ductility reduction. Introducing high-density coherent nano-dispersoids into the W matrix is a highly efficient strategy to break the tradeoff of the strength-ductility performance. In this work, we performed helium (He) ion irradiation on coherent oxide-dispersoids strengthened (ODS) W to investigate the effect of coherent nanoparticle interfaces on the behavior of He bubbles. The results show that the diameter and density of He bubbles in ODS W are close to that in W at low dose of He ion irradiation. The radiation-induced hardening increment of ODS W, being 25% lower than that of pure W, suggests the involvement of the coherent interface in weakening He ion irradiation-induced hardening and emphasizes the potential of coherent nano-dispersoids in enhancing the radiation resistance of W-based materials.

A First-principles analysis of Structural, Electronic, Elastic, and vacant properties of BCC, FCC and HCP tungsten at different electronic temperatures

Since tungsten (W) was considered as the most promising plasma facing materials (PFMs) in fusion reactors, there has been extensive research on the physical performance of W-PFMs. It is found that under the extreme conditions in a fusion reactor, W-PFMs should be in a nonequilibrium state of high electronic temperature and low ionic temperature. This leads to the possibility of non-thermal phase transitions, where the crystal structure of the tungsten material may change from body-centered cubic (bcc) phase to hexagonal close-packed (hcp) phase or face-centered cubic (fcc) phase. Consequently, it is necessary to investigate the relevant physical properties of hcp-W and fcc-W under the electron-excited state. In this work, the fundamental physical properties, including atomic structures, electronic structures, elastic constants, and vacancy formation energies, of bcc-W, hcp-W and fcc-W, were theoretically calculated at various electronic temperatures. The mechanical stability of these three phases was also systematically analyzed under varying electronic temperatures. The results of this research are expected to provide a certain guidance in the optimization of W-PFMs in future fusion reactors.

Characterization of W-Er2O3 alloy prepared by co-deposition method and spark plasma sintering

Oxide dispersion strengthened tungsten (ODS-W) alloys are promising plasma facing material in future fusion reactors. In this study, composite powders of W-Er2O3 were prepared by co-deposition method and hydrogen reduction, and then consolidated by spark plasma sintering (SPS). Microstructure evolution during co-deposition process was investigated using X-ray diffraction (XRD), scanning electron microscope (SEM), transmission electron microscope (TEM) and Fourier transform infrared spectrometer (FTIR). Also, simulation using DMol3 module in Materials Studio was carried out to understand the co-deposition mechanism. The results showed that H+ of W-OH bond in white tungstic acid was replaced by Er3+ due to coordination reaction during co-deposition process, leading to formation of ring-like chelate of (WO)nEr. This indicates that doping of Er3+ was achieved at a molecular level. Consequently, Er2O3 particles were relatively uniformly distributed in W-Er2O3 alloy, and the hardness and compressive strength reached 627 ± 19 HV and 1640 ± 40 MPa, respectively.

In situ melting phenomena on W plasma-facing components for lower divertor during long-pulse plasma operations in EAST

Tungsten (W) is one of the most promising plasma-facing materials for future fusion devices. Although its melting point is the highest among all metals, it still has great risk of melting under extremely high plasma heat fluxes, which is a big concern for ITER and future reactors. Actively cooled W plasma-facing components (PFCs) with both monoblocks and flat-type structure have been successfully installed in the lower divertor of the EAST tokamak since 2021, and provide a good opportunity for direct comparison of the damage mechanism for the two types of PFCs. Various in situ melting phenomena on the lower divertor have been observed by CCD cameras, which have been further verified by post-mortem inspections. Severe melting and even exfoliation of the edge-beveled W plates were observed on some W/Cu flat-type components at horizontal outer targets. Many droplets were ejected during long-pulse operations, which induced a significant increase of W impurities and total irradiation in the core plasma, and thus greatly deteriorated the plasma performance and even caused disruptions. Two different shaping structures of flat-type PFCs show different positions of melting and the corresponding mechanisms. Slight melting was found on the sharp leading edges of W/Cu monoblocks between cassette modules (inter-CM) for horizontal targets with small droplet ejection, which was much improved compared to that observed on the upper W divertor, illustrating that the application of a large-sized bevel chamfer inter-CM was generally effective. In addition, an unexpected melting phenomenon on the dome plate was attributed to the extreme transient heat flux during disruption with runaway electrons. The application of both types of W/Cu PFCs for the divertor provides important experiences and lessons for the engineering design and optimization of divertor PFCs in future fusion devices.

Irradiation hardening of WK-based PFM exposed to 14 MeV protons

W-based alloys are perceived as one of the most promising plasma-facing materials (PFMs). During the service of fusion reactors, W-based PFMs will be exposed to 14 MeV fusion neutrons, causing significant displacement damage and transmutation, consequently, the irradiation-induced defects will seriously deteriorate its service performance. The irradiation effect is always a critical issue in fusion reactors and concerns the performance and reliability of W-based PFMs. In this work, WK-0.05Ti alloy was irradiated under 14MeV protons, and notable irradiation hardening was observed using microhardness and nanoindentation characterization. Then, the mechanism of W-Re precipitate formation induced by K-bubble was revealed using transmission electron microscopy. Finally, the variation of hardness values at different irradiation depths was explained based on theoretical calculations and experimental data. The present work elucidates the effect of K-bubbles on the evolution of irradiation defects in WK-based PFMs during irradiation and reveals the underlying mechanisms contributing to the irradiation hardening at different depths. Moreover, it is expected to provide additional insights into the R&D of robust plasma facing components based on WK-based materials.

The evolution of hydrogen bubbles during ion irradiation and annealing in molybdenum for neutral beam injection inductively coupled plasma source

Molybdenum (Mo) is considered as a promising plasma-facing material, which is expected to be used in the manufacture of critical components for the neutral beam injection (NBI) inductively coupled plasma (ICP) source. Hydrogen isotope retention and bubble formation are regarded as the unavoidable issues that degrade the metal properties in the ion source. In this work, the evolution of hydrogen bubbles in Mo pre-irradiated by 35 eV hydrogen ion energy during different annealing processes is studied through conductive atomic force microscopy (CAFM) observation. Meanwhile, the corresponding numerical simulation is applied to analyze the mechanisms of hydrogen bubble growth. It is found that the average size and density of hydrogen bubbles are enhanced with the increase of hydrogen ion fluence at 400 K in experiments and simulations. In the subsequent annealing process, it is demonstrated that annealing temperature has a certain effect on bubble growth and evolution. No obvious changes in bubbles are observed at the annealing temperature below 573 K. The average radius of hydrogen bubbles is increased by ∼1.0-2.5 nm in Mo annealed at 673-873 K. The bubble size increases with the rise of annealing temperature, indicating that the thermal annealing promotes the growth of bubbles. When the temperature rises above 873 K, the change in bubble size is not obvious. The simulation results indicate that temperature-induced changes in the internal pressure of hydrogen bubbles leads to the growth of bubbles, but this mechanism is not significant under high annealing temperatures due to the release of hydrogen from the surface.

Microstructure and mechanical property evolution of He-implanted nanochannel W film under post-annealing

Tungsten (W) has been selected as the most promising Plasma Facing materials (PMFs) due to its excellent properties. One of the most challenged problems for W is the He retention. Nowaday, a nanochannel structure film was presented to improve the He irradiation resistance of W. In this work, the microstructure and mechanical property evolution of He-implanted nanochannel W film under post-annealing from 573 to 1273 K were studied. A significant growth of He bubbles and the formation of large size holes at grain boundaries (GBs) were observed after post-annealing at 1273 K. In addition, a formation model of pores structure on surface caused by high energy He implantation was proposed. Moreover, a mitigated lattice swelling was observed at 1273 K due to the He-V clusters and He bubbles migration and coarsening. The shape of the He bubbles at 1273 K was also investigated in detail. The nanochannel film has a better resistance to the degradation of Young's modulus under the He implantation than compact film, due to the loose nanochannel. Hardness increment induced by the implantation of He was analyzed. The hardening effect of the He-implanted film annealing at 1073 K, resulting from the transition of He-V clusters or He bubbles distributions from dislocations to distributed vacancies, should be notice. Finally, the holes at GBs led to the decrease of both Young's modulus and hardness.

Recovery and recrystallization of warm-rolled tungsten during helium thermal desorption spectroscopy annealing

Tungsten has been perceived as one of the most promising plasma-facing materials for future fusion reactors. Its behavior under helium irradiation has been extensively studied. In this work, we focus on the recovery and recrystallization of helium-implanted surface grains in warm-rolled polycrystalline tungsten. Mechanically polished specimens with and without 40 keV helium implantation were heated to 1650 °C during helium thermal desorption spectroscopy experiment. Our results suggest that recovery and recrystallization occurred via subgrain coarsening and removal of low-angle grain boundaries in as-polished specimen. Its average grain size increased from 1.70 to 8.89 μm, and its fraction of low-angle grain boundary decreased from 0.77 to 0.29. By contrast, little changes of grain size and grain boundary misorientation angle were found in helium-implanted surface grains. Such enhanced thermal stability is attributed to the pinning effect of helium bubbles on grain boundaries, and the underlying mechanism is discussed based on the proposed model of cellular microstructure.

Atomic Simulations of the Interaction between a Dislocation Loop and Vacancy-Type Defects in Tungsten

Tungsten (W) is considered to be the most promising plasma-facing material in fusion reactors. During their service, severe irradiation conditions create plenty of point defects in W, which can significantly degrade their performance. In this work, we first employ the molecular static simulations to investigate the interaction between a 1/2[111] dislocation loop and a vacancy-type defect including a vacancy, di-vacancy, and vacancy cluster in W. The distributions of the binding energies of a 1/2[111] interstitial and vacancy dislocation loop to a vacancy along different directions at 0 K are obtained, which are validated by using the elasticity theory. The calculated distributions of the binding energies of a 1/2[111] interstitial dislocation loop to a di-vacancy and a vacancy cluster, showing a similar behavior to the case of a vacancy. Furthermore, we use the molecular dynamics simulation to study the effect of a vacancy cluster on the mobility of the 1/2[111] interstitial dislocation loop. The interaction is closely related to the temperature and their relative positions. A vacancy cluster can attract the 1/2[111] interstitial dislocation loop and pin it at low temperatures. At high temperatures, the 1/2[111] interstitial dislocation loop can move randomly. These results will help us to understand the essence of the interaction behaviors between the dislocation loop and a vacancy-type defect and provide necessary parameters for mesoscopic scale simulations.

Effect of He on the hardness and incipient plasticity of W films during nanoindentation

In future fusion reactors, the production and accumulation of helium (He) will lead to a dramatic mechanical failure for the material. Tungsten (W) is considered as one of the most promising plasma facing materials. In order to study the effect of He on mechanical properties of W, the novel He-charged W films with a homogeneous He dose profile were fabricated by the magnetron sputtering device. The prepared samples in this study can well avoid the problems like damage gradient effect, indentation size effect, and soft substrate effect during the nanoindentation test, which will appear in the samples obtained by conventional high energy He+ implantation. Nanoindentation results showed that the hardness of the films increased with the increasing He content, and the hardness-depth relationship is in good agreement with the Nix-Gao model. The pop-in events can be observed in the pure W film, which were absent in the He-charged films. Results of molecular dynamics (MD) simulations proved that, the He bubble not only promotes the nucleation and propagation of dislocations, but also hinders the dislocation movement, which leads to a smoother transition from elastic to plastic deformation in the indentation process.

Deuterium retention and surface morphology modification in oxidized tungsten exposed to D plasma

Tungsten (W) is the most promising plasma facing material in future fusion reactor. However, oxidation occurs readily on W and might influence the hydrogen isotopes retention behavior in W. To explore the mechanism of oxidation on the hydrogen isotopes retention behavior in W, oxidized W samples have been exposed to D plasma (32.3 eV, 2 × 1020 Dm−2s−1) in a linear plasma device. Surface morphology and deuterium (D) retention behavior in W have been characterized after exposure to D plasma with fluence of 1023 D/m2. After D plasma exposure, a porous layer has been found on the surface of oxidized W. The main chemical content of the porous layer is pure W. The existence of pre-oxide layer contributes the increase of retention amount of D in the form of D2O and may contribute to the decrease of the retention amount of D in the form of HD/D2.

Atomic-Scale Hidden Point-Defect Complexes Induce Ultrahigh-Irradiation Hardening in Tungsten.

Tungsten displays high strength in extreme temperature and radiation environments and is considered a promising plasma facing material for fusion nuclear reactors. Unlike other metals, it experiences substantial irradiation hardening, which limits service life and presents safety concerns. The origin of ultrahigh-irradiation hardening in tungsten cannot be well-explained by conventional strengthening theories. Here, we demonstrate that irradiation leads to near 3-fold increases in strength, while the usual defects that are generated only contribute less than one-third of the hardening. An analysis of the distribution of tagged atom-helium ions reveals that more than 87% of vacancies and helium atoms are unaccounted for. A large fraction of helium-vacancy complexes are frozen in the lattice due to high vacancy migration energies. Through a combination of in situ nanomechanical tests and atomistic calculations, we provide evidence that irradiation hardening mainly originates from high densities of atomic-scale hidden point-defect complexes.

Laser melting deposited self-passivating 90W-10Cr coatings on reduced activation ferritic/martensitic steel using 90W-7Ni-3Fe interlayers

• 90W-7Ni-3Fe interlayers were used for the joining of RAFM steel and W-Cr coatings by LMD. • The interlayer with high dilution rate is effective in relieving residual stress. • W-Cr coatings have reliable anti-oxidation performance compared to pure W. As promising plasma facing materials considering the loss of coolant accident (LOCA) in fusion reactors, W-Cr coatings need to be joined to the structural components manufactured with reduced activation ferritic/martensitic (RAFM) steel. Laser melting deposition (LMD) is a promising additive manufacturing technique to deposit W-Cr coatings on the RAFM steels. In order to reduce the stress concentration at the coating/substrate interface, 90W-7Ni-3Fe interlayers were deposited on the RAFM steel by LMD before the deposition of W-Cr coatings. Two interlayers with similar thickness of about 750 μm were prepared with different dilution rates from the RAFM steel substrates. Both interlayers have improved the joining of RAFM steels and W-Cr coatings by eliminations of cracks, which originate at the substrate/coating interfaces without interlayers. No pores or cracks are observed at the substrate/interlayer interfaces, or the interlayer/coating interfaces. The top and mid-top regions of coatings have unique microstructures showing W particles embedded in Cr based matrix. Compared to the high hardness at the substrate/interlayer and interlayer/coating interfaces for the high dilution rate interlayer, the interface hardness is much lower for the low dilution rate interlayer. Moreover, sharp decreases of hardness happen at the interfaces for the low dilution rate interlayer. Pores are formed at the W-Cr coatings due to the lack of fusion, residual of delivered gas, or evaporations of low melting point elements. For W-Cr coatings deposited on the low dilution rate interlayers, cracks can be observed. The thermogravimetric analysis under air atmosphere indicates reliable anti-oxidation performance of the W-Cr coatings compared to the pure W.

Helium effects on the surface and subsurface evolutions in single-crystalline tungsten

Tungsten (W) has been perceived as one of the most promising plasma facing materials (PFMs) for future fusion reactors. In the past decade, its behavior under irradiation and helium (He) plasma interaction has been extensively studied. However, some key knowledge gaps still exist, such as the influence of crystallographic orientation on the surface and subsurface evolutions. In this work, we focus on the He ion-beam irradiation damage effects in mirror-polished single-crystalline W samples with three different surface planes of {100}, {110} and {111}. Irradiation was performed at room temperature using 40 keV He+ to a fluence of 1 × 1016/cm2, followed by thermal desorption spectroscopy (TDS) up to ~1920 K. The microstructures of He-irradiated W before and after TDS heat treatment were characterized by scanning and transmission electron microscopy. Subsurface He bubbles were imaged in all irradiated samples, but newly formed <111>-oriented surface grains and surface blisters were only observed in W {100} and {110} starting orientations. These results reveal that radiation damage, He thermal desorption, and surface/subsurface evolution are all strongly dependent on crystallographic orientation. Underlying physical mechanisms are discussed based on ion channeling effects, He-vacancy interactions, and surface diffusion. These findings provide new insights into He effects in W.

Tungsten-chromium coatings on reduced activation ferritic/martensitic steels prepared by laser melting deposition process

Tungsten-chromium alloys are promising plasma facing materials in future fusion reactors considering the loss of coolant accident (LOCA). A laser melting deposition (LMD) additive manufacturing technique has been used in this study to deposit 90W-10Cr coatings on reduced activation ferritic/martensitic (RAFM) steels. The results show that the coatings have metallurgical bonding with the RAFM steel substrates. The W-steel joint have graded transitions of compositions, microstructures, and phases along the deposition direction due to the dilution effects of the substrates, which are beneficial to reduce thermal stresses of W-steel joints. Pores and cracks are seen in the coatings prepared using different process parameters. Low and high input energy densities can both lead to pore formations due to the combined effects of lack of fusion, residual of delivered gas, and evaporations of low melting point elements. Cracking is driven by the internal stress during LMD process and liquation of binding matrix. Fe7W6 intermetallics are mainly formed near the bonding boundaries of W-steel joint, showing a dispersed distribution either scattered in the matrix or surrounding the scattered W particles. The top and mid-top regions of coatings have unique microstructures showing W particles embedded in Cr-Fe based matrix, which are supposed to have reliable performance both during the operations of reactors and the LOCA.

Hardening effect of multi-energyW2+-ion irradiation on tungsten–potassium alloy**Project supported by the National Natural Science Foundation of China (Grant Nos. 11975160 and 11775149). One of the authors, Kun Zhang, was supported by the Fundamental Research Funds for the Central Universities, China.

Tungsten is one of the most promising plasma-facing materials (PFMs) to be used in the nuclear fusion reactor as divertor material in the future. In this work, W2+-ions bombardment is used to simulate the neutron irradiation damage to commercial pure tungsten (W) and rolled tungsten–potassium (W–K). The 7 MeV of 3 × 1015 W2+-ions/cm2, 3 MeV of 4.5 × 1014 W2+, and 2 MeV of 3 × 1014 W2+-ions/cm2 are applied at 923 K in sequence to produce a uniform region of 100 nm–400 nm beneath the sample surface with the maximum damage value of 11.5 dpa. Nanoindentation is used to inspect the changes in hardness and elastic modulus after self-ion irradiation. Irradiation hardening occurred in both materials. The irradiation hardening of rolled W–K is affected by two factors: one is the absorption of vacancies and interstitial atoms by potassium bubbles, and the other is the interaction between potassium bubbles and dislocations. Under the condition of 11.5 dpa, the capability of defect absorption can reach a threshold. As a result, dislocations finally dominate the hardening of rolled W–K. Specific features of dislocation loops in W–K are further observed by transmission electron microscopy (TEM) to explain the hardening effect. This work might provide valuable enlightenment for W–K alloy as a promising plasma facing material candidate.

Thermal shock resistance of tungsten with various deformation degrees under transient high heat flux

Tungsten (W) is considered as the most promising plasma facing material in fusion device, which will be exposed to steady and transient heat loads. Usually, the deformation degree of W influences its thermal shock properties. So we evaluated the thermal shock resistance of rolled pure tungsten (PW) with 60%, 90% deformation degrees and W-1.0wt%La2O3 (WL10) with 52%, 88% deformation degrees using electron beam at an absorbed power density of 0.22 GW m−2 for 5 ms and further discussed the relationship between the thermal shock resistance and microstructures, thermal-mechanical properties. The absorbed beam current (30 mA), electron beam acceleration voltage (120 kV) and loaded area (4 × 4 mm2) is used to estimate the absorbed power density. The results indicated that PW-90%, LW-88% exhibited smaller grain size, higher relative density, microhardness but lower strength, thermal conductivity compared to PW-60%, LW-52%. The cracking threshold was <0.22 GW m−2 for PW-60%, LW-52% and >0.22 GW m−2 for PW-90%, LW-88%. Elliptic left pores in PW-60% and LW-52% aggravated the effect of stress concentration cracking during the transient high heat flux test and decreased the cracking threshold altough they exhibited high strength and thermal conductivity.