Microstructure Evolution in ZrCx with Different Stoichiometries Irradiated by Four MeV Au Ions

Abstract

ZrCx ceramics with different stoichiometries were irradiated under a four MeV Au ion beam in doses of 2 × 1016 ions/cm2 at room temperature, corresponding to ~130 dpa. Grazing incidence, X-ray diffraction and transmission electron microscopy were performed to study the radiation damage and microstructure evolution in ZrCx ceramics. With the decrease in C/Zr ratio, the expansion of ZrCx lattice became smaller after irradiation. Some long dislocation lines formed at the near-surface, while, in the area with the greatest damage (depth of ~400 nm), large amounts of dislocation loops formed in ZrC, ZrC0.9 and ZrC0.8. With the increase in carbon vacancy concentration, the size of the dislocation loops gradually decreased. Few dislocation loops were found in ZrC0.7 after irradiation, and only black-dot defects were found in the area with the greatest damage. For the non-stoichiometric ZrCx, with the increase of the intrinsic vacancies, the number of C interstitials caused by irradiation decreased, and the recombination barrier of C Frenkel pairs reduced. The above factors will reduce the total number of C interstitials after cascade cooling, suppressing the formation and growth of dislocation loops, which is significant for the enhancement of the tolerance of radiation damage.

Synthesis, mechanical, and thermophysical properties of high‐entropy (Zr,Ti,Nb,Ta,Hf)C0.8 ceramic

“…Due to the promoted diffusion, more un‐synthesized single‐phase high‐entropy TMCs could be obtained by introducing carbon vacancies. It is also promising that the irradiation tolerance could be enhanced significantly in high‐entropy TMCs with moderate carbon vacancies 31,32 . Although several studies have reported the synthesis and preliminary mechanical properties of high‐entropy TMCs, the detailed influence of carbon stoichiometry on microstructure, mechanical properties, thermal properties, and oxidation resistances has not been reported.…”

“…It is also promising that the irradiation tolerance could be enhanced significantly in high-entropy TMCs with moderate carbon vacancies. 31,32 Although several studies have reported the synthesis and preliminary mechanical properties of high-entropy TMCs, the detailed influence of carbon stoichiometry on microstructure, mechanical properties, thermal properties, and oxidation resistances has not been reported. Therefore, the aim of the present work is to make a detailed comparison of the sinterability, mechanical properties, and thermophysical properties between the well-studied (Zr,Ti,Nb,Ta,Hf)C and nonstoichiometric (Zr,Ti,Nb,Ta,Hf)C 0.8 and evaluate their applications .…”

Reduced He ion irradiation damage in ZrC-based high-entropy ceramics

“…In addition, the charge compensation is typical in HECs when elements with different valence states are mixed. Vacancies can be formed to maintain valence balance [32], which can restrain the lattice expansion during irradiation [18,33]. Up to the present, there are very few related studies on the irradiation behavior of HECs, particularly high-entropy carbide ceramics.…”

First‐principles study of the thermal properties of Zr2C and Zr2CO

“…ZrC is a candidate to replace SiC in TRISO fuel particles due to its high melting point, high thermal conductivity at very high temperatures, and low neutron absorption cross‐sections 2 . It has come to light that ZrC, especially ZrC x containing significant fractions of carbon vacancies, is chemically and mechanically stable under irradiation and when exposed to fission products 3–4 . In addition, using ZrC in TRISO particles could increase the gas exit temperature of VHTRs, and increase the thermal efficiency as a result 5 .…”

“…2 It has come to light that ZrC, especially ZrC x containing significant fractions of carbon vacancies, is chemically and mechanically stable under irradiation and when exposed to fission products. [3][4] In addition, using ZrC in TRISO particles could increase the gas exit temperature of VHTRs, and increase the thermal efficiency as a result. 5 The thermal stability of ZrC x can also enhance the accident-tolerance of nuclear fuels.…”