During manufacturing and operation, different kinds of defects, e.g., delamination or surface cracks, may be generated in the plasma-facing components (PFCs) of a Tokamak device. To ensure the safety of the PFCs, vari...During manufacturing and operation, different kinds of defects, e.g., delamination or surface cracks, may be generated in the plasma-facing components (PFCs) of a Tokamak device. To ensure the safety of the PFCs, various kinds of nondestructive testing (NDT) techniques are needed for different defect and failure mode. This paper gives a review of the recently developed ultrasonic testing (UT) and laser thermography methods for inspection of the delamination and surface cracks in PFCs. For monoblock W/Cu PFCs of divertor, the bonding quality at both W-Cu and Cu- CuCrZr interfaces was qualified by using UT with a focus probe during manufacturing. A noncontact, coupling-free and flexible ultrasonic scanning testing system with use of an electromagnetic acoustic transducer and a robotic inspection manipulator was introduced then for the in-vessel inspection of delamination defect in first wall (FW). A laser infrared thermography testing method is highlighted for the on-line inspection of delamination defect in FW through the vacuum vessel window of the Tokamak reactor. Finally, a new laser spot thermography method using laser spot array source was described for the online inspection of the surface cracks in FW.展开更多
In a fusion reactor, plasma-facing components（PFCs） will suffer severe thermal shock; behavior and performance of PFCs under high heat flux（HHF） loads are of major importance for the long-term stable operation of ...In a fusion reactor, plasma-facing components（PFCs） will suffer severe thermal shock; behavior and performance of PFCs under high heat flux（HHF） loads are of major importance for the long-term stable operation of the reactor. This work investigates the thermo-mechanical behaviors of tungsten armor under high heat loads by the method of finite element modeling and simulating. The temperature distribution and corresponding thermal stress changing rule under different HHF are analyzed and deduced. The Manson–Coffin equation is employed to evaluate the fatigue lifetime（cyclic times of HHF loading） of W-armored first wall under cyclic HHF load. The results are useful for the formulation design and structural optimization of tungsten-armored PFCs for the future demonstration fusion reactor and China fusion experimental thermal reactor.展开更多
W/Cu Functionally Graded Materials (FGM) was designed not only for reducing the thermal stress caused by the mismatch of thermal expansion coefficients, but also for combining the features of W, Mo - high plasma-erosi...W/Cu Functionally Graded Materials (FGM) was designed not only for reducing the thermal stress caused by the mismatch of thermal expansion coefficients, but also for combining the features of W, Mo - high plasma-erosion resistance and the advantages of Cu - high heat conductivity and ductility. Four different fabrication processes for W/Cu or Mo/Cu, including hot-pressing, Cu infiltration of sintered porosity-graded W skeleton, spark plasma sintering and plasma spraying, were investigated and compared. It was foundthat the hot-pressing process is difficult to keep the designed composition gradient, while the other three processes are successful in making W/Cu or Mo/Cu FGM. Meanwhile, microstructures and composition gradients are analyzed with SEM and EDAX.展开更多
Classical molecular dynamics has been used to study the interactions between tung- sten （W） plasma-facing materials （PFMs） and dust grains. The impact velocity of dust grains is in the range from 324 m/s to 3240 m...Classical molecular dynamics has been used to study the interactions between tung- sten （W） plasma-facing materials （PFMs） and dust grains. The impact velocity of dust grains is in the range from 324 m/s to 3240 m/s. The main effect of dust grains with low impact velocity is deposition. However, a material surface can be damaged by high velocity dust grains. The cumulative damage of impacting dust grains has also been take into account. When the impact velocity is low, no significant damage is detected but a porous firm forms on the surface. Serious damage can be produced on PFMs if the impact velocity is high.展开更多
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...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 tempera-ture 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.展开更多
This contribution summarized the recent studies of tungsten-based plasma-facing materials in the linear plasma device like the simulator for tokamak edge plasma(STEP),focusing on the examination of newly developed tun...This contribution summarized the recent studies of tungsten-based plasma-facing materials in the linear plasma device like the simulator for tokamak edge plasma(STEP),focusing on the examination of newly developed tungsten(W)-based materi-als and plasma-induced defects in pure W.Pure W,W-V,W-Y_(2)O_(3)and W-ZrC samples were exposed to a high-flux plasma of~1021-1022 m^(−2)s^(−1) with a fluence up to 1026 m^(−2) at a surface temperature below 500 K.The investigation of fundamental evolution of plasma-induced defects in pure W indicated a critical role of hydrogen-dislocation interactions.Suppressed surface blistering was observed in all W-based materials,but deuterium desorption behavior and retention were distinct with respect to different materials.The studies showed that the linear plasma device like the STEP was indispensable in the understanding of plasma-material interactions and the qualification of new materials for future fusion reactors.展开更多
Tungsten-potassium(potassium-doped tungsten or WK),initially known from the electric filament industry,is a promising plasma-facing material(PFM)in future fusion facilities like International Thermonuclear Experimenta...Tungsten-potassium(potassium-doped tungsten or WK),initially known from the electric filament industry,is a promising plasma-facing material(PFM)in future fusion facilities like International Thermonuclear Experimental Reactor(ITER).However,the brittle nature of W and irradiation-induced defects of WK materials may result in a risk of deuterium-tritium reaction failure in fusion reactors.Previous studies revealed that advanced W with ultrafine grains and nanostructures might be able to address these problems.However,K-doped W,a rapidly developed material for PFMs,lacks a systematical sum-mary.In this review,we firstly describe the powder metallurgy and plastic deformation for the preparation of WK.Then,the mechanical properties of WK and thermal shock resistance results are reviewed.Important issues such as irradiation damages from neutron,heavy ion,and plasma(H isotope or He)irradiation are also discussed.Hitherto,WK under irradia-tions shows comparable or even better performances compared with other counterparts such as ITER grade pure tungsten.This review could be benefitial to the future efforts of improving the ductility and irradiation tolerance of WK materials.展开更多
The thermal conductivity of plasma-facing materials(PFM)exposed to intense radiation is a critical concern for the reliable usage of materials in fusion reactors.However,limited research has been performed regarding t...The thermal conductivity of plasma-facing materials(PFM)exposed to intense radiation is a critical concern for the reliable usage of materials in fusion reactors.However,limited research has been performed regarding the thermal conductivity of structures that rapidly change in a short time during collision cascade processes under irradiation.In this study,we employed the tight-binding(TB)method to investigate the electronic thermal conductivity(κ_(e))of tungsten-based systems during various cascading processes.We found thatκ_(e) values sharply decrease within the initial 0.3 picoseconds and then partially recover at a slow pace;this is closely linked to the evolution of defects and microstructural distortions.The increase in the initial kinetic energy of the primary knock-on atom and the presence of a high concentration of hydrogen atoms further decrease theκ_(e) values.Conversely,higher temperatures have a significant positive effect onκ_(e).Furthermore,the presence of a grain boundary∑5[001](130)substantially reducesκ_(e),whereas the absorption effect of point defects by the grain boundary has little influence onκ_(e) during cascades.Our findings provide a theoretical basis for evaluating changes in the thermal conductivity performance of PFMs during their usage in nuclear fusion reactors.展开更多
W-based WTaVCr refractory high entropy alloys (RHEA) may be novel and promising candidate materials for plasma facing components in the first wall and diverter in fusion reactors. This alloy has been developed by a po...W-based WTaVCr refractory high entropy alloys (RHEA) may be novel and promising candidate materials for plasma facing components in the first wall and diverter in fusion reactors. This alloy has been developed by a powder metallurgy process combining mechanical alloying and spark plasma sintering (SPS). The SPSed samples contained two phases, in which the matrix is RHEA with a body-centered cubic structure, while the oxide phase was most likely Ta2VO6through a combined analysis of X-ray diffraction (XRD),energy-dispersive spectroscopy (EDS), and selected area electron diffraction (SAED). The higher oxygen affinity of Ta and V may explain the preferential formation of their oxide phases based on thermodynamic calculations. Electron backscatter diffraction (EBSD) revealed an average grain size of 6.2μm. WTaVCr RHEA showed a peak compressive strength of 2997 MPa at room temperature and much higher micro-and nano-hardness than W and other W-based RHEAs in the literature. Their high Rockwell hardness can be retained to at least 1000°C.展开更多
Oxide-dispersion-strengthened tungsten(ODS-W)and a Cu Cr Zr alloy were bonded by a three-step process:(i)surface nano-activation,(ii)copper plating followed by annealing,and(iii)diffusion bonding.The morphological and...Oxide-dispersion-strengthened tungsten(ODS-W)and a Cu Cr Zr alloy were bonded by a three-step process:(i)surface nano-activation,(ii)copper plating followed by annealing,and(iii)diffusion bonding.The morphological and structural evolutions of ODS-W and the interface of the ODS-W/CuCrZr joint during these processes have been thoroughly studied by X-ray diffraction,scanning electron microscopy,energy dispersive spectrometry,and high-resolution transmission electron microscopy.After surface nanoactivation,a nanoporous structure of ODS-W with an average pore size of~100 nm was obtained,and the Y_(2)O_(3)particles therein remained unchanged.A Cu coating was tightly bonded with the surface nanoactivated ODS-W after Cu plating and annealing.An interaction layer embedded with nanosized W particles was formed at the interface between ODS-W and plated Cu after the three-step process.Consequently,well-cohesive ODS-W/Cu and ODS-W/Y_(2)O_(3)/Cu interfaces were formed.The ODS-W/Cu Cr Zr joint showed high shear strengths(up to 201 MPa)and effective bonded area ratios(>98%).The developed three-step bonding process between ODS-W and the Cu Cr Zr alloy provides an effective support for future plasma-facing components in nuclear fusion reactor applications.展开更多
In this work, a time-of-flight （TOF） mass spectrometer has been used to investigate the distribution of intermediate species and formation process of carbon clusters. The graphite sample was ablated by Nd：YAG laser...In this work, a time-of-flight （TOF） mass spectrometer has been used to investigate the distribution of intermediate species and formation process of carbon clusters. The graphite sample was ablated by Nd：YAG laser （532 nm and 1064 nm）. The results indicate that the maximum size distribution shifted towards small cluster ions as the laser fluence increased, which happened because of the fragmentation of larger clusters in the hot plume. The temporal evolution of ions was measured by varying the delay time of the ion extraction pulse with respect to the laser irradiation, which was used to provide distribution information of the species in the ablated plasma plume. When the laser fluence decreased, the yield of all of the clusters obviously dropped.展开更多
Plasma-facing components in thermonuclear reactors primarily consist of plasma-facing materials and heat-sink materials.Tungsten-based materials are currently regarded as the most promising candidates as plasma-facing...Plasma-facing components in thermonuclear reactors primarily consist of plasma-facing materials and heat-sink materials.Tungsten-based materials are currently regarded as the most promising candidates as plasma-facing materials,while Cu alloys are typically utilized as heat-sink materials.However,bonding tungsten-based materials and Cu alloys together is challenging due to the inherent immiscibility of W and Cu.This review outlines advanced bonding technologies for tungstenbased materials and Cu alloys by tailoring joint interfaces.These technologies encompass:(i)direct diffusion bonding of W and Cu using high-temperature conditions(close to the melting point of Cu)structure,with an emphasis on elucidating the underlying thermodynamic mechanisms through the construction of thermodynamic models and molecular dynamics simulations;(ii)combined technologies involving surface treatments of tungsten-based materials,copper embedding,and diffusion bonding,along with an analysis of the mechanisms that enhance joint properties through tailored interface structures.The review also provides insights into future research directions for bonding between tungsten-based materials and Cu alloys.These advancements may offer significant support for plasma-facing components in future thermonuclear fusion reactors.展开更多
As a plasma-facing material(PFM),tungsten is subjected to the synergistic irradiation of steady-state plasma and transient thermal load.The study investigated the irradiation damage and subsequent performance degradat...As a plasma-facing material(PFM),tungsten is subjected to the synergistic irradiation of steady-state plasma and transient thermal load.The study investigated the irradiation damage and subsequent performance degradation of tungsten(W)at different preheating temperatures under synergistic irradiation of 180 MW·m^(-2).The findings revealed that increasing the preheating temperature resulted in the fusion of smaller grains into larger ones,a decrease in dislocation line density,and a reduction in reflection.When the preheating temperature was below the ductile-to-brittle transition temperature(DBTT,~400℃),defects trapped hydrogen and formed a bubble-like structure on the surface with crack widths ranging from 100 to 300 nm.The pinning effect of hydrogen retention caused an increase in hardness,but as the pinning effect weakened,large cracks led to a decrease in hardness.Preheating W above the DBTT resulted in no bubble-like structures on the surface and reduced crack widths to below 100 nm,with a corresponding increase in hardness,indicating reduced damage.Therefore,when tungsten is used as a PFM in nuclear fusion reactions,its preheating temperature should be controlled above the DBTT to ensure it remains in a ductile state with strong irradiation resistance.展开更多
As the interface closest to the edge plasma of a fusion reactor reacting deuterium(D)and tritium(T),plasma-facing material(PFM)need to withstand extreme service conditions with the high particle flux,high heat load,an...As the interface closest to the edge plasma of a fusion reactor reacting deuterium(D)and tritium(T),plasma-facing material(PFM)need to withstand extreme service conditions with the high particle flux,high heat load,and neutrons with energy up to 14.1 MeV.Tungsten(W)is the primary candidate of PFMs in future fusion reactors due to its high melting point,good thermal conductivity,excellent irradiation resistance,and low hydrogen/helium retention.So far,powder metallurgy is a leading route for the preparation of W-based PFM.An alternative approach could be the coating technique,which has advantages on fabricating W PFMs and plasma-facing component(PFC)simultaneously.In the past several years,inspiring results were achieved in the preparation process and performance evaluation of the W coating with high purity,excellent thermal conductivity,and thickness at the millimeter level by atmospheric pressure chemical vapor deposition(APCVD).No obvious grain growth and hardness decrease were observed when the annealing temperature was lower than 1500°C,indicating its good thermal stability.The as-deposited coating exhibited a comparable thermal shock resistance with the conventional W bulk.While the polished sample showed a high crack threshold(0.33-0.44 GW·m^(−2))when exposed to edge localized mode like transient at room temperature,compared to the unpolished counterpart.Irradiation performance of the chemical vapor deposition(CVD)-W exposed to deuterium(D)plasma and fission neutron were also evaluated.Additionally,the practicality of preparation of large-scale W-based PFM by this technique is also demonstrated.This paper gives a short overview on the recent research and development status of the thick W coating prepared by APCVD at Xiamen Tungsten Co.,Ltd and Southwestern Institute of Physics for using as PFM and PFC in fusion devices.展开更多
Nuclear fusion has enormous potential to greatly affect global energy production. The next-generation tokamak ITER, which is aimed at demonstrating the feasibility of energy production from fusion on a commercial scal...Nuclear fusion has enormous potential to greatly affect global energy production. The next-generation tokamak ITER, which is aimed at demonstrating the feasibility of energy production from fusion on a commercial scale, is under construction. Wall erosion, material transport, and fuel retention are known factors that shorten the lifetime of ITER during tokamak operation and give rise to safety issues. These factors, which must be understood and solved early in the process of fusion reactor design and development, are among the most important concerns for the community of plasma-wall interaction researchers. To date, laser techniques are among the most promising methods that can solve these open ITER issues, and laser-induced breakdown spectroscopy (LIBS) is an ideal candidate for online monitoring of the walls of current and next-generation (such as ITER) fusion devices. LIBS is a widely used technique for various applications. It has been considered recently as a promising tool for analyzing plasma-facing components in fusion devices in situ. This article reviews the experiments that have been performed by many research groups to assess the feasibility of LIBS for this purpose.展开更多
The change in surface damage/microstructures and its effects on the hydrogen(H)isotope/helium(He)dynamic behavior are the key factors for investigating issues of tungsten(W)-based plasma-facing materials(PFMs)in fusio...The change in surface damage/microstructures and its effects on the hydrogen(H)isotope/helium(He)dynamic behavior are the key factors for investigating issues of tungsten(W)-based plasma-facing materials(PFMs)in fusion such as surface erosion,H/He retention and tritium(T)inventory.Complex surface damage/microstructures are introduced in W by high-temperature plasma irradiation and new material design,typically including pre-damage and multi-ion co-deposition induced structures,solute elements and related composites,native defects like dislocations and interfaces,and nanostructures.Systematic experimental and theoretical researches were performed on H isotope/He retention in complex W-based materials in the past decades.In this review,we aim to provide an overview of typical surface damage/microstructures and their effects on H/He retention in W,both in the experiment and multiscale modeling.The distribution/state,dynamics evolution,and interaction with defects/microstructures of H/He are generally summarized at different scales.Finally,the current difficulties,challenges and future directions are also discussed about H/He retention in complex W-based PFMs.展开更多
Considering that tungsten(W)materials served as the plasma-facing material in the fusion reactor would be exposed to edge-localized modes(ELMs)-like thermal shock loading accompanied with He-ion irradiation,the W-TiC ...Considering that tungsten(W)materials served as the plasma-facing material in the fusion reactor would be exposed to edge-localized modes(ELMs)-like thermal shock loading accompanied with He-ion irradiation,the W-TiC composite produced with a wet-chemical method was conducted by the dual effects from the laser beam thermal shock first and He-ion irradia-tion later in this work.The microstructure changes of the W-TiC composite before and after two tests were characterized by scanning electron microscopy or transmission electron microscopy.After the laser beam thermal shock test,there was an obvious interface on the exposed surface of the W-TiC composite.Several main cracks and melting areas could be found nearby the interface and center,respectively.Furthermore,a mixture of tungsten oxide and TiC was easy to aggregate and form into circle areas surrounding the melting area.The thermal shock tested that W-TiC composite was then subjected to the He-ion irradiation.The typical features of fuzz structures could be detected on the surface of the W-TiC composite apart from the center of the melting area.Notably,several nano-sized He bubbles deeply distributed at grain boundaries in the melting area,owing to the grain boundary functioning as the free path for He diffusion.展开更多
Tungsten-chromium-yttrium(WCrY)smart alloys are foreseen as the first wall material for future fusion devices such as Demonstration Power Plant(DEMO).While suppressing W oxidation during accidental conditions,they sho...Tungsten-chromium-yttrium(WCrY)smart alloys are foreseen as the first wall material for future fusion devices such as Demonstration Power Plant(DEMO).While suppressing W oxidation during accidental conditions,they should behave like pure W during plasma operation due to preferential sputtering of the lighter alloying elements Cr,Y,and W enrichment of the surface.In this paper,the erosion performance of WCrY and W samples simultaneously exposed to deuterium(D)plasma with the addition of 1%of the projectile ions being argon(Ar)ions at an ion energy of 120 eV is compared.With reference to the previous experiments at 120 eV in pure D plasma,the erosion for both WCrY and W is enhanced by a factor of~7.Adding Ar to the D plasma suppresses significant W enrichment previously found for pure D plasma.To investigate the impact of the plasma exposure onto the oxidation performance,plasma-exposed and non-exposed reference samples were oxidised in a dry atmosphere.Results show,on the one hand,that the oxida-tion suppression of WCrY in comparison to pure W is preserved during the plasma performance.On the other hand,it becomes evident that edge effects imposed by the geometry of the samples used in plasma experiments play a significant role for the oxidation behaviour.展开更多
基金the National Magnetic Confinement Fusion Program of China(Grant 2013GB113005)the National Natural Science Foundation of China(Grants51577139 and 11502192)for funding
文摘During manufacturing and operation, different kinds of defects, e.g., delamination or surface cracks, may be generated in the plasma-facing components (PFCs) of a Tokamak device. To ensure the safety of the PFCs, various kinds of nondestructive testing (NDT) techniques are needed for different defect and failure mode. This paper gives a review of the recently developed ultrasonic testing (UT) and laser thermography methods for inspection of the delamination and surface cracks in PFCs. For monoblock W/Cu PFCs of divertor, the bonding quality at both W-Cu and Cu- CuCrZr interfaces was qualified by using UT with a focus probe during manufacturing. A noncontact, coupling-free and flexible ultrasonic scanning testing system with use of an electromagnetic acoustic transducer and a robotic inspection manipulator was introduced then for the in-vessel inspection of delamination defect in first wall (FW). A laser infrared thermography testing method is highlighted for the on-line inspection of delamination defect in FW through the vacuum vessel window of the Tokamak reactor. Finally, a new laser spot thermography method using laser spot array source was described for the online inspection of the surface cracks in FW.
基金the financial supports from the ITER-National Magnetic Confinement Fusion Program (Nos. 2014 GB123000 and 2010 GB109000)the National Natural Science Foundation of China (No. 51172016)
文摘In a fusion reactor, plasma-facing components（PFCs） will suffer severe thermal shock; behavior and performance of PFCs under high heat flux（HHF） loads are of major importance for the long-term stable operation of the reactor. This work investigates the thermo-mechanical behaviors of tungsten armor under high heat loads by the method of finite element modeling and simulating. The temperature distribution and corresponding thermal stress changing rule under different HHF are analyzed and deduced. The Manson–Coffin equation is employed to evaluate the fatigue lifetime（cyclic times of HHF loading） of W-armored first wall under cyclic HHF load. The results are useful for the formulation design and structural optimization of tungsten-armored PFCs for the future demonstration fusion reactor and China fusion experimental thermal reactor.
文摘W/Cu Functionally Graded Materials (FGM) was designed not only for reducing the thermal stress caused by the mismatch of thermal expansion coefficients, but also for combining the features of W, Mo - high plasma-erosion resistance and the advantages of Cu - high heat conductivity and ductility. Four different fabrication processes for W/Cu or Mo/Cu, including hot-pressing, Cu infiltration of sintered porosity-graded W skeleton, spark plasma sintering and plasma spraying, were investigated and compared. It was foundthat the hot-pressing process is difficult to keep the designed composition gradient, while the other three processes are successful in making W/Cu or Mo/Cu FGM. Meanwhile, microstructures and composition gradients are analyzed with SEM and EDAX.
基金supported by National Natural Science Foundation of China(No.11075186)National Magnetic Confinement Fusion Science Program of China(No.2013GB107004)
文摘Classical molecular dynamics has been used to study the interactions between tung- sten （W） plasma-facing materials （PFMs） and dust grains. The impact velocity of dust grains is in the range from 324 m/s to 3240 m/s. The main effect of dust grains with low impact velocity is deposition. However, a material surface can be damaged by high velocity dust grains. The cumulative damage of impacting dust grains has also been take into account. When the impact velocity is low, no significant damage is detected but a porous firm forms on the surface. Serious damage can be produced on PFMs if the impact velocity is high.
基金the National Natural Science Foundation of China(Grant Nos.51771184,11735015,11575241,51801203 and 11575231)the Natural Science Foundation of Anhui Province(Grant No.1808085QE132)the Open Project of State Key Laboratory of Environment Friendly Energy Materials(Grant No.18kfhg02).
文摘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 tempera-ture 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.
基金the National Nature Science Foundation of China(Grant 51720105006 and 11805007)the Science and Technology on Surface Physics and Chemistry Laboratory(Grant 02020317).
文摘This contribution summarized the recent studies of tungsten-based plasma-facing materials in the linear plasma device like the simulator for tokamak edge plasma(STEP),focusing on the examination of newly developed tungsten(W)-based materi-als and plasma-induced defects in pure W.Pure W,W-V,W-Y_(2)O_(3)and W-ZrC samples were exposed to a high-flux plasma of~1021-1022 m^(−2)s^(−1) with a fluence up to 1026 m^(−2) at a surface temperature below 500 K.The investigation of fundamental evolution of plasma-induced defects in pure W indicated a critical role of hydrogen-dislocation interactions.Suppressed surface blistering was observed in all W-based materials,but deuterium desorption behavior and retention were distinct with respect to different materials.The studies showed that the linear plasma device like the STEP was indispensable in the understanding of plasma-material interactions and the qualification of new materials for future fusion reactors.
基金the National Natural Science Foundation of China(Grant Nos.11775149 and 11475118).
文摘Tungsten-potassium(potassium-doped tungsten or WK),initially known from the electric filament industry,is a promising plasma-facing material(PFM)in future fusion facilities like International Thermonuclear Experimental Reactor(ITER).However,the brittle nature of W and irradiation-induced defects of WK materials may result in a risk of deuterium-tritium reaction failure in fusion reactors.Previous studies revealed that advanced W with ultrafine grains and nanostructures might be able to address these problems.However,K-doped W,a rapidly developed material for PFMs,lacks a systematical sum-mary.In this review,we firstly describe the powder metallurgy and plastic deformation for the preparation of WK.Then,the mechanical properties of WK and thermal shock resistance results are reviewed.Important issues such as irradiation damages from neutron,heavy ion,and plasma(H isotope or He)irradiation are also discussed.Hitherto,WK under irradia-tions shows comparable or even better performances compared with other counterparts such as ITER grade pure tungsten.This review could be benefitial to the future efforts of improving the ductility and irradiation tolerance of WK materials.
基金supported by the Collaborative Innovation Program of Hefei Science Center of CAS(No.2022HSC-CIP007)。
文摘The thermal conductivity of plasma-facing materials(PFM)exposed to intense radiation is a critical concern for the reliable usage of materials in fusion reactors.However,limited research has been performed regarding the thermal conductivity of structures that rapidly change in a short time during collision cascade processes under irradiation.In this study,we employed the tight-binding(TB)method to investigate the electronic thermal conductivity(κ_(e))of tungsten-based systems during various cascading processes.We found thatκ_(e) values sharply decrease within the initial 0.3 picoseconds and then partially recover at a slow pace;this is closely linked to the evolution of defects and microstructural distortions.The increase in the initial kinetic energy of the primary knock-on atom and the presence of a high concentration of hydrogen atoms further decrease theκ_(e) values.Conversely,higher temperatures have a significant positive effect onκ_(e).Furthermore,the presence of a grain boundary∑5[001](130)substantially reducesκ_(e),whereas the absorption effect of point defects by the grain boundary has little influence onκ_(e) during cascades.Our findings provide a theoretical basis for evaluating changes in the thermal conductivity performance of PFMs during their usage in nuclear fusion reactors.
基金supported by the National Science Foundation under Grant No.CMMI-1762190The research was performed in part in the Nebraska Nanoscale Facility:National Nanotechnology Coordinated Infrastructure and the Nebraska Center for Materials and Nanoscience (and/or NERCF),which are supported by the National Science Foundation under Award ECCS:2025298+1 种基金the Nebraska Research Initiativesupported by the U.S.Department of Energy,Office of Nuclear Energy under DOE Idaho Operations Office Contract DE-AC07-051D14517 as part of a Nuclear Science User Facilities experiment。
文摘W-based WTaVCr refractory high entropy alloys (RHEA) may be novel and promising candidate materials for plasma facing components in the first wall and diverter in fusion reactors. This alloy has been developed by a powder metallurgy process combining mechanical alloying and spark plasma sintering (SPS). The SPSed samples contained two phases, in which the matrix is RHEA with a body-centered cubic structure, while the oxide phase was most likely Ta2VO6through a combined analysis of X-ray diffraction (XRD),energy-dispersive spectroscopy (EDS), and selected area electron diffraction (SAED). The higher oxygen affinity of Ta and V may explain the preferential formation of their oxide phases based on thermodynamic calculations. Electron backscatter diffraction (EBSD) revealed an average grain size of 6.2μm. WTaVCr RHEA showed a peak compressive strength of 2997 MPa at room temperature and much higher micro-and nano-hardness than W and other W-based RHEAs in the literature. Their high Rockwell hardness can be retained to at least 1000°C.
基金supported by the National Key Research and Development Program of China(No.2017YFE0302600 and No.2018YFB0703904)。
文摘Oxide-dispersion-strengthened tungsten(ODS-W)and a Cu Cr Zr alloy were bonded by a three-step process:(i)surface nano-activation,(ii)copper plating followed by annealing,and(iii)diffusion bonding.The morphological and structural evolutions of ODS-W and the interface of the ODS-W/CuCrZr joint during these processes have been thoroughly studied by X-ray diffraction,scanning electron microscopy,energy dispersive spectrometry,and high-resolution transmission electron microscopy.After surface nanoactivation,a nanoporous structure of ODS-W with an average pore size of~100 nm was obtained,and the Y_(2)O_(3)particles therein remained unchanged.A Cu coating was tightly bonded with the surface nanoactivated ODS-W after Cu plating and annealing.An interaction layer embedded with nanosized W particles was formed at the interface between ODS-W and plated Cu after the three-step process.Consequently,well-cohesive ODS-W/Cu and ODS-W/Y_(2)O_(3)/Cu interfaces were formed.The ODS-W/Cu Cr Zr joint showed high shear strengths(up to 201 MPa)and effective bonded area ratios(>98%).The developed three-step bonding process between ODS-W and the Cu Cr Zr alloy provides an effective support for future plasma-facing components in nuclear fusion reactor applications.
基金supported by the National Magnetic Confinement Fusion Science Program of China(No.2013GB109005)National Natural Science Foundation of China(No.11175035)+2 种基金Chinesisch-Deutsches Forschungs Project(GZ768)the Fundamental Research Funds for the Central Universities of China(Nos.DUT12ZD(G)01,DUT14ZD(G)04)MMLab Research Project(DP1051208)
文摘In this work, a time-of-flight （TOF） mass spectrometer has been used to investigate the distribution of intermediate species and formation process of carbon clusters. The graphite sample was ablated by Nd：YAG laser （532 nm and 1064 nm）. The results indicate that the maximum size distribution shifted towards small cluster ions as the laser fluence increased, which happened because of the fragmentation of larger clusters in the hot plume. The temporal evolution of ions was measured by varying the delay time of the ion extraction pulse with respect to the laser irradiation, which was used to provide distribution information of the species in the ablated plasma plume. When the laser fluence decreased, the yield of all of the clusters obviously dropped.
基金supported by the National Natural Science Foundation of China(No.51971153)the National Key Research and Development Program of China(No.2017YFE0302600)+1 种基金Scientific Research Initial Funding of Taiyuan University of Science and Technology(No.20232088)the Award Fund for Outstanding Doctors in Shanxi Province(No.20242005)。
文摘Plasma-facing components in thermonuclear reactors primarily consist of plasma-facing materials and heat-sink materials.Tungsten-based materials are currently regarded as the most promising candidates as plasma-facing materials,while Cu alloys are typically utilized as heat-sink materials.However,bonding tungsten-based materials and Cu alloys together is challenging due to the inherent immiscibility of W and Cu.This review outlines advanced bonding technologies for tungstenbased materials and Cu alloys by tailoring joint interfaces.These technologies encompass:(i)direct diffusion bonding of W and Cu using high-temperature conditions(close to the melting point of Cu)structure,with an emphasis on elucidating the underlying thermodynamic mechanisms through the construction of thermodynamic models and molecular dynamics simulations;(ii)combined technologies involving surface treatments of tungsten-based materials,copper embedding,and diffusion bonding,along with an analysis of the mechanisms that enhance joint properties through tailored interface structures.The review also provides insights into future research directions for bonding between tungsten-based materials and Cu alloys.These advancements may offer significant support for plasma-facing components in future thermonuclear fusion reactors.
基金funded by the National Key Research and Development Program of China(No.22022YFE03030003)。
文摘As a plasma-facing material(PFM),tungsten is subjected to the synergistic irradiation of steady-state plasma and transient thermal load.The study investigated the irradiation damage and subsequent performance degradation of tungsten(W)at different preheating temperatures under synergistic irradiation of 180 MW·m^(-2).The findings revealed that increasing the preheating temperature resulted in the fusion of smaller grains into larger ones,a decrease in dislocation line density,and a reduction in reflection.When the preheating temperature was below the ductile-to-brittle transition temperature(DBTT,~400℃),defects trapped hydrogen and formed a bubble-like structure on the surface with crack widths ranging from 100 to 300 nm.The pinning effect of hydrogen retention caused an increase in hardness,but as the pinning effect weakened,large cracks led to a decrease in hardness.Preheating W above the DBTT resulted in no bubble-like structures on the surface and reduced crack widths to below 100 nm,with a corresponding increase in hardness,indicating reduced damage.Therefore,when tungsten is used as a PFM in nuclear fusion reactions,its preheating temperature should be controlled above the DBTT to ensure it remains in a ductile state with strong irradiation resistance.
基金This work is financially supported by the National Magnetic Confinement Fusion Program of China(Grant No.2018YFE0312100)the National Natural Science Foundation of China(Grant Nos.11975092 and 11905045).
文摘As the interface closest to the edge plasma of a fusion reactor reacting deuterium(D)and tritium(T),plasma-facing material(PFM)need to withstand extreme service conditions with the high particle flux,high heat load,and neutrons with energy up to 14.1 MeV.Tungsten(W)is the primary candidate of PFMs in future fusion reactors due to its high melting point,good thermal conductivity,excellent irradiation resistance,and low hydrogen/helium retention.So far,powder metallurgy is a leading route for the preparation of W-based PFM.An alternative approach could be the coating technique,which has advantages on fabricating W PFMs and plasma-facing component(PFC)simultaneously.In the past several years,inspiring results were achieved in the preparation process and performance evaluation of the W coating with high purity,excellent thermal conductivity,and thickness at the millimeter level by atmospheric pressure chemical vapor deposition(APCVD).No obvious grain growth and hardness decrease were observed when the annealing temperature was lower than 1500°C,indicating its good thermal stability.The as-deposited coating exhibited a comparable thermal shock resistance with the conventional W bulk.While the polished sample showed a high crack threshold(0.33-0.44 GW·m^(−2))when exposed to edge localized mode like transient at room temperature,compared to the unpolished counterpart.Irradiation performance of the chemical vapor deposition(CVD)-W exposed to deuterium(D)plasma and fission neutron were also evaluated.Additionally,the practicality of preparation of large-scale W-based PFM by this technique is also demonstrated.This paper gives a short overview on the recent research and development status of the thick W coating prepared by APCVD at Xiamen Tungsten Co.,Ltd and Southwestern Institute of Physics for using as PFM and PFC in fusion devices.
文摘Nuclear fusion has enormous potential to greatly affect global energy production. The next-generation tokamak ITER, which is aimed at demonstrating the feasibility of energy production from fusion on a commercial scale, is under construction. Wall erosion, material transport, and fuel retention are known factors that shorten the lifetime of ITER during tokamak operation and give rise to safety issues. These factors, which must be understood and solved early in the process of fusion reactor design and development, are among the most important concerns for the community of plasma-wall interaction researchers. To date, laser techniques are among the most promising methods that can solve these open ITER issues, and laser-induced breakdown spectroscopy (LIBS) is an ideal candidate for online monitoring of the walls of current and next-generation (such as ITER) fusion devices. LIBS is a widely used technique for various applications. It has been considered recently as a promising tool for analyzing plasma-facing components in fusion devices in situ. This article reviews the experiments that have been performed by many research groups to assess the feasibility of LIBS for this purpose.
基金This work was financially supported by the National Natural Science Foundation of China(Grant Nos.11975018,11775254 and 11534012)the National Magnetic Confinement Fusion Energy Research Project(Grant No.2018YEF0308100)+2 种基金the Science Challenge Project(Grant No.TZ2018004)the Youth Innovation Promotion Association of Chinese Academy of Sciences(CAS)(Grant No.2016386)Director Grants of Hefei Institutes of Physics Science,Chinese Academy of Sciences(CASHIPS).
文摘The change in surface damage/microstructures and its effects on the hydrogen(H)isotope/helium(He)dynamic behavior are the key factors for investigating issues of tungsten(W)-based plasma-facing materials(PFMs)in fusion such as surface erosion,H/He retention and tritium(T)inventory.Complex surface damage/microstructures are introduced in W by high-temperature plasma irradiation and new material design,typically including pre-damage and multi-ion co-deposition induced structures,solute elements and related composites,native defects like dislocations and interfaces,and nanostructures.Systematic experimental and theoretical researches were performed on H isotope/He retention in complex W-based materials in the past decades.In this review,we aim to provide an overview of typical surface damage/microstructures and their effects on H/He retention in W,both in the experiment and multiscale modeling.The distribution/state,dynamics evolution,and interaction with defects/microstructures of H/He are generally summarized at different scales.Finally,the current difficulties,challenges and future directions are also discussed about H/He retention in complex W-based PFMs.
基金the National Natural Science Foundation of China(Grant No.51574101)the Fundamental Research Funds for the Central Universities(Grant Nos.PA2018GDQT0010,PA2019GDZC0096,JZ2019HGTA0040)+2 种基金the Foundation of Laboratory of Nonferrous Metal Material and Processing Engineering of Anhui Province(15CZS08031)the Natural Science Foundation of Anhui Province(Grant Nos.201904b11020034,1908085ME115)the Foundation of Laboratory of Nonferrous Metal Material and Processing Engineering of Anhui Province,the Open Foundation of Key Laboratory of Advanced Functional Materials,Devices of Anhui Province and Double First Class enhancing independent innovation and social service capabilities of Hefei University of Technology(Grant No.45000-411104/011).
文摘Considering that tungsten(W)materials served as the plasma-facing material in the fusion reactor would be exposed to edge-localized modes(ELMs)-like thermal shock loading accompanied with He-ion irradiation,the W-TiC composite produced with a wet-chemical method was conducted by the dual effects from the laser beam thermal shock first and He-ion irradia-tion later in this work.The microstructure changes of the W-TiC composite before and after two tests were characterized by scanning electron microscopy or transmission electron microscopy.After the laser beam thermal shock test,there was an obvious interface on the exposed surface of the W-TiC composite.Several main cracks and melting areas could be found nearby the interface and center,respectively.Furthermore,a mixture of tungsten oxide and TiC was easy to aggregate and form into circle areas surrounding the melting area.The thermal shock tested that W-TiC composite was then subjected to the He-ion irradiation.The typical features of fuzz structures could be detected on the surface of the W-TiC composite apart from the center of the melting area.Notably,several nano-sized He bubbles deeply distributed at grain boundaries in the melting area,owing to the grain boundary functioning as the free path for He diffusion.
基金the framework of the EURO fusion Consortium,the Euratom research and training programme 2014-2018 and 2019-2020(Grant 633053)the European Commission through the Erasmus Mundus International Doctoral College in Fusion Science and Engineering(FUSION-DC).
文摘Tungsten-chromium-yttrium(WCrY)smart alloys are foreseen as the first wall material for future fusion devices such as Demonstration Power Plant(DEMO).While suppressing W oxidation during accidental conditions,they should behave like pure W during plasma operation due to preferential sputtering of the lighter alloying elements Cr,Y,and W enrichment of the surface.In this paper,the erosion performance of WCrY and W samples simultaneously exposed to deuterium(D)plasma with the addition of 1%of the projectile ions being argon(Ar)ions at an ion energy of 120 eV is compared.With reference to the previous experiments at 120 eV in pure D plasma,the erosion for both WCrY and W is enhanced by a factor of~7.Adding Ar to the D plasma suppresses significant W enrichment previously found for pure D plasma.To investigate the impact of the plasma exposure onto the oxidation performance,plasma-exposed and non-exposed reference samples were oxidised in a dry atmosphere.Results show,on the one hand,that the oxida-tion suppression of WCrY in comparison to pure W is preserved during the plasma performance.On the other hand,it becomes evident that edge effects imposed by the geometry of the samples used in plasma experiments play a significant role for the oxidation behaviour.
