【摘 要】
：
W-3wt%Lu2O3 composites were prepared by mechanical milling and spark plasma sintering.The W-3wt%Lu2O3 composites were maintained in a vacuum at room temperature,and irradiated with a He+(He+ion beam f
【机 构】
：
School of Materials Science and Engineering,Hefei University of Technology,Hefei 230009,China
【出 处】
：
第13届中日双边先进能源系统和聚变裂变工程材料会议(CIS-13)
论文部分内容阅读
W-3wt%Lu2O3 composites were prepared by mechanical milling and spark plasma sintering.The W-3wt%Lu2O3 composites were maintained in a vacuum at room temperature,and irradiated with a He+(He+ion beam flux,1×1022 ions/m2·s)ion beam(beam energy,50 eV and 80 eV)for 2 h using the Large-Power Materials Irradiation Experiment System.X-ray Diffraction(XRD),Field-emission scanning electron microscopy,and high-resolution transmission electron microscopy were used to characterize the surface morphology of composites after exposure to a low-energy,high-flux He+irradiation.Results showed that relative density of the W-Lu2O3 composites was 96.3%.
其他文献
Ferritic/martensitic(F/M)steel is regarded as one of the most promising candidate structural materials for GEN IV,fusion reactor and accelerator driven system(ADS),which have been widely studied in th
A novel model with built-in tungsten filament on the basis of join W/Cu functionally graded layer was proposal and investigated.Using the ABAQUS software on the basis of finite element method,the dist
Helium(He)behavior in metals has been a subject of both scientific interest and technological significance for decades.Because of its extremely low solubility,high mobility and self-trapping character
In the design of future fusion reactors,tungsten(W)is currently a leading material choice for plasma facing materials(PFMs)for divertor.For plasma facing components(PFCs)with relatively small particle
In the High temperature gas cooled reactor(HTGR),the radio nuclides produced by fission and activating reactions would be carried into circulation by helium.Tritium is one of the most important radioa
Tritium breeder materials are significant for blanket design of fusion reactor.However,during blanket operation,the ceramic breeder materials will be subjected to neutron irradiation which could be de
Vanadium alloy is considered as the first candidate material for self-cooled lithium blanket for its excellent neutron resistance、mechanical property and compatibility with lithium.In the real applica
W-ZrC composites reinforced with Sc2O3 particles were produced through powder metallurgy(PM)and subsequent spark plasma sintering(SPS)at 1700 °C and 13.7 KN.The microstructure and properties of the co
Reduced activation ferritic steel alloys such as F82H are the candidate materials for the first wall of DEMO reactors.For self-cooled breeding blankets,the first wall will be subjected to bi-direction
The transient thermal shock performances and relevant damage mechanism of pure W and lutecium doped W alloys were investigated via a repetitive laser beam heat load test to simulate the transient heat