Effects of helium irradiation dose and temperature on the damage evolution of Ti3SiC2 ceramic

The effects of 400 keV helium ion irradiation dose and temperature on the microstructure of the Ti3SiC2 ceramic were systematically investigated by grazing incidence x-ray diffraction, scanning electron microscopy, and transmission electron microscopy.The helium irradiation experiments were performed at both room temperature(RT) and 500℃ with a fluence up to 2.0 × 1017 He+/cm2 that resulted in a maximum damage of 9.6 displacements per atom.Our results demonstrate that He irradiations produce a large number of nanometer defects in Ti3SiC2 lattice and then cause the dissociation of Ti3SiC2 to TiC nano-grains with the increasing He fluence.Irradiation induced cell volume swelling of Ti3SiC2 at RT is slightly higher than that at 500℃, suggesting that Ti3SiC2 is more suitable for use in a high temperature environment.The temperature dependence of cell parameter evolution and the aggregation of He bubbles in Ti3SiC2 are different from those in Ti3AlC2.The formation of defects and He bubbles at the projected depth would induce the degradation of mechanical performance.

Microstructure and hardening effect of pure tungsten and ZrO2 strengthened tungsten under carbon ion irradiation at 700℃

Microstructure evolution and hardening effect of pure tungsten and W-1.5%ZrO_(2) alloy under carbon ion irradiation are investigated by using transmission electron microscopy and nano-indentation.Carbon ion irradiation is performed at 700℃ with irradiation damages ranging from 0.25 dpa to 2.0 dpa.The results show that the irradiation defect clusters are mainly in the form of dislocation loop.The size and density of dislocation loops increase with irradiation damages intensifying.The W-1.5%ZrO_(2) alloy has a smaller dislocation loop size than that of pure tungsten.It is proposed that the phase boundaries have the ability to absorb and annihilate defects and the addition of ZrO_(2) phase improves the sink strength for irradiation defects.It is confirmed that the W-1.5% ZrO_(2) alloy shows a smaller change in hardness than the pure tungsten after being irradiated.From the above results,we conclude that the addition of ZrO_(2) into tungsten can significantly reduce the accumulation of irradiated defects and improve the irradiation resistance behaviors of the tungsten materials.

Preliminary assessment of high-entropy alloys for tritium storage

Tritium is the key fuel in nuclear fusion reactors.With the development of the international thermonuclear experimental reactor(ITER)project,the annual requirement of tritium has increased up to several kilograms.The candidate materials for tritium storage have many shortcomings such as insufficient kinetic performance,disproportionation effect,poor oxidation resistance,and poor helium(He)retaining ability.Therefore,it is urgent to develop a novel material system which satisfies all the requirements of tritium storage materials.High-entropy alloys(HEAs)have a unique structure of severe lattice distortion and have attracted much attention as hydrogen storage materials due to their high storing capacity and great hydrogenation performance.The distorted lattice helps to provide more interstitial sites for accommodating H atoms and enhance the He retaining ability by slowing down the He diffusion in the HEA lattice.In this work,the current research status of tritium storage materials,including the background and the basic criterion of tritium storage materials,as well as the disadvantages of the current materials,has been reviewed.Moreover,the theoretical and experimental studies of HEAs,focusing on the hydrogenation properties and the defect evolution in the distorted lattice,have been summarized.The HEAs may have great potential as tritium storage materials due to their potential hydrogenation performance and He retaining ability.Finally,the existing challenges and future development directions are also proposed.