Fabrication And Irradiation Resistance Of ZrC/ZTA Ceramic Composites

Active,safe and orderly develop nuclear energy is a strategic choice for China to achieve the goal of“Double carbon”,and guarantee sustainable,high-quality economic and social development.Material is a major constraint to the development of nuclear energy.The Generation IV nuclear energy systems urgently need the support of materials with high strength,high temperature resistance,irradiation resistance,and corrosion resistance.Ceramics have become candidate materials for new high temperature reactors owing to excellent physical and chemical properties.However,there are some problems such as poor toughness and lack of irradiation data for single-phase nuclear ceramics.Therefore,it is significant to develop new toughened ceramic composites and evaluate their radiation resistance for the screening of nuclear ceramic materials.Zr C/ZTA ceramic composites synthesized by Spark Plasma Sintering were selected for our research.Improving the comprehensive thermomechanical properties by optimizing the composition,powder size,and sintering temperature.Irradiation experiments for optimized Zr C/ZTA ceramic composites were conducted with helium ions and carbon ions.Irradiation damage effects mainly include microstructure damage,mechanical properties,and its mechanisms.1 Optimized fabrication of toughened Zr C/ZTA ceramic composites（1）Zr C/ZTA ceramic composites exhibit great chemical compatibility during sintering process.Combining the microstructure and mechanical properties,the Zr C/ZTA ceramic composites display excellent comprehensive mechanical properties at 25 vol%of Zr C.（2）Two particle sizes of Zr C（50 nm and 1-3μm）were used for preparing Zr C/ZTA ceramic composites.Due to the excellent microstructure,μm-Zr C/ZTA composites exhibit superior properties to nm-Zr C/ZTA including the relative density,Vickers hardness,flexural strength,fracture toughness,and thermal conductivity.（3）The average grain size of Zr C（25 vol%）/ZTA ceramic composites increased with sintering temperatures（1500,1600,1700,1800℃）,and there were abnormally grown grains at 1800℃.The relative density,Vickers hardness,fracture toughness,and thermal conductivity were optimal at the sintering temperature of 1700℃.Overall,the optimized Zr C/ZTA ceramic composites for the irradiation damage study were prepared with a sintering temperature of 1700℃,a grain size of 1-3μm in original Zr C powder,and a Zr C content of 25 vol%,which exhibit refined microstructure and excellent mechanical properties.2 Irradiation response of toughened Zr C/ZTA ceramic compositesIrradiation experiments for optimized Zr C/ZTA ceramic composites were conducted with helium ions(500 ke V,He²⁺)and carbon ions(1 Me V,C⁴⁺).Irradiation parameters are as follows:RT（1.0～20.0）×10¹⁶ He²⁺/cm~2,RT（4.0～10.0）×10¹⁷ C⁴⁺/cm~2,650℃1.0×10¹⁷ C⁴⁺/cm~2.The irradiation damage effect was studied by grazing incidence X-ray（GIXRD）,transmission electron microscopy（TEM）,and nanoindentation.The results showed that the phases consist ofα-Al₂O₃,t-Zr O₂,and Zr C,and no phase transformation occurred.（1）Helium ion irradiation:The degradation of crystallinity and structure as well as lattice swelling after irradiation.In damage peak region,the size and density of helium bubbles increase significantly.Helium bubbles are mainly distributed inα-Al₂O₃,a few inside t-Zr O₂ grains,and no significant helium bubble aggregation in Zr C.A large number of helium bubbles aggregate at grain and phase boundaries.At the influence of 2.0×10¹⁷ He²⁺/cm~2,a few micro-cracks formed in someα-Al₂O₃ and t-Zr O₂grains in the peak damage region,parallel to the sample surface.The softening at the influence of 2.0×10¹⁷ He²⁺/cm~2 may be related to the formation of micro-cracks,which is one of the energy dissipation mechanisms.（2）Carbon ion irradiation:The defect morphology,size,and density show difference in each phase with irradiation damage increasing.Inα-Al₂O₃ phase,the dislocation loops are mainly in the shape of“coffee bean”with a small size.At room temperature,the size and density of dislocation loops are related to the damage,with large defect size and high density in peak damaged area.At high temperature（650℃）,only the defects density increases with the irradiation fluences.In t-Zr O₂ phase,defects are composed of dislocation lines,dislocation loops,and defect clusters mixed.No significant differences were found in defect morphology,size and distribution under different damages.Another characteristic of defects in Zr C is the stacking fault,the density of stacking fault correlates with irradiation damage.Nano-hardness and elastic modulus show slight hardening with irradiation.