Abstract
Combining the advantages of a wet chemical method and spark plasma sintering, carbide-doped materials W-1wt%TiC and W-1wt%ZrC were prepared. Microstructural evolution in W-1wt%TiC and W-1wt%ZrC under irradiation of 5 keV He+ at 600 °C to fluences up to 5.0 × 1021 ions/m2 with ion flux of about 8.8 × 1017 ions/m2s was investigated by transmission electron microscopy (TEM). The dislocation loop number density of W-1wt%TiC was higher than that of W-1wt%ZrC, but the average loop size of the W-1wt%TiC was in average smaller. There were no observable helium bubbles in W-1wt%TiC and W-1wt%ZrC, exhibiting higher radiation resistance to He+ compared to pure W. He+ pre-damaged and undamaged W-1wt%TiC and W-1wt%ZrC samples were irradiated by 5 keV D2+ to estimate the D retention in doped W materials. The irradiation damage impact of He+ on deuterium retention was examined by a method of thermal desorption spectroscopy (TDS). Compared with the undamaged samples, it was illustrated that D2 retention of W-1wt%TiC and W-1wt%ZrC increased after He+ pre-irradiation.
Sign-in/Register to access full text options 
Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have
Schedule a callDisclaimer: All third-party content on this website/platform is and will remain the property of their respective owners and is provided on "as is" basis without any warranties, express or implied. Use of third-party content does not indicate any affiliation, sponsorship with or endorsement by them. Any references to third-party content is to identify the corresponding services and shall be considered fair use under The CopyrightLaw.
