Effect of dislocation on helium retention in deformed 316L stainless steel

来源 :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) | 被引量 : 0次 | 上传用户:amies
下载到本地 , 更方便阅读
声明 : 本文档内容版权归属内容提供方 , 如果您对本文有版权争议 , 可与客服联系进行内容授权或下架
论文部分内容阅读
  Austenitic stainless steels 316L have been widely utilized for industrial application,especially,as the candidate structural materials for fusion reactor.It is well known that helium atoms can be trapped by several type defects,such as dislocations and vacancies,and would be formed to bubbles.In the present work,the effect of dislocation on helium retention in 316L stainless steel was investigated.In which the dislocations were introduced by cold rolling machine,and helium were produced by 50 keV He+irradiation at room temperature.Positron annihilation spectroscopy methods were performed to characterize the formation and behavior of defects and helium defect clusters.The positron annihilation lifetime and S parameters with positron energy will be used to demonstrate the detailed information and interaction of helium atoms with dislocation trapping sites.At the same time,the desorption energies of helium defect clusters were estimated from the thermal desorption spectroscopy.It shows that high density dislocation have stronger inhibitory effect for helium desorption at 800-1200 K.
其他文献
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
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