Preparation And Environmental Damage Behavior Of ZrC_x Ceramics With High Tolerance Of Radiation Damage

Due to the combination of high temperature,high neutron does and extremely corrosive environment of Generation IV nuclear reactor systems,the development of new nuclear materials with good radiation resistance,corrosion resistance and high thermostability is imminent.ZrC is considered as the TRISO-coating fuel particles,fuel cladding or inert matrix materials due to its high melting point,high thermal conductivity,low neutron absorption cross section and excellent resistance to the attack by fission products.In the present work,ZrCx ceramics with various C/Zr ratios of 0.6–1.0 were fabricated by two-step reactive hot pressing.The effects of C/Zr ratio on densification behavior,microstructure,mechanical properties,thermal properties,irradiation damage behavior under 4MeV Au ions?2×10¹66 ions/cm²?and corrosion behavior in high temperature water vapor of ZrCx ceramic have been investigated systematically.Futhermore,the mechanisms of irradiation damage under Au ions and corrosion in high temperature water were proposed.The densification temperature of ZrCx ceramics decreases and lattice parameter of ZrCx decreases linearly with the decrease of C/Zr ratio.The grain size of ZrCx ceramics increases with the increase of sintering temperature and the decrease of C/Zr ratio.A cubic Zr2C-type vacancy ordered phase forms in ZrC0.6.With the increase of sintering temperature and the decrease of C/Zr ratio,the comprehensive mechanical properties of ZrCx ceramics decrease.ZrC0.9 ceramics sintered at 2000°C possesses good comprehensive properties such as elastic modulus of 390 GPa,Vickers hardness of 17.9GPa,flexural strength of 358 MPa and indentation fracture toughness of 3.8 MPa·m^1/2.The measured thermal conductivity ZrCx ceramics increase monotonously with the increases of temperature.With decreasing the C/Zr ratio,the thermal conductivity,specific heat and thermal diffusion coefficient decrease due to the increased scattering of conducting phonons and electron by carbon vacancies.The ZrC after irradiation can be divided into five layers according to the defect structure characteristics in the depth direction.Due to the strong ionization of Au ions,the surface of ZrC after irradiation forms a¹5nm thick ZrO2 nanocrystalline layer?layer I?;In the non-elastic collision region,the atomic temperature is higher,and the dislocations are easy to slip in the region,forming a layer with lower density dislocation line?layer II?.There have largest damage dose in Layer III.High temperature of inelastic collision makes the dislocation loops to slip and grow easily,eventually forming a layer with larger size dislocation loops?layer III?.The average damage dose in layer IV is lower than that in layer III.While the dislocation loop can formed.However,due to the lower temperature of the atom,it is not conducive to the slip and growth of the dislocation loop,thus forming a layer with smaller size dislocation loops?layer IV?.The layer V corresponds to the lowest damage dose,thus it is diffcult to form dislocation loops.The Frenkel defect exists in the form of"black spot"defects?layer V?.The intrinsic C vacancies can inhibit the formation and grow of dislocation loops,resulting from the number of C interstitial under irradiation decreasing,the the Frenkel pairs recombination energy of C reducing,and total C vacancy concentration increasing,all with the intrinsic C vacancy concentration increasing.ZrCx ceramics show excellent radiation resistance under Au ion irradiation at a dose of about 130dpa.With the increase of C vacancies,the radiation resistance of ZrCx ceramics have been improved.The lattice expansion of ZrC0.6 ceramic after irradiation is only 0.013%,and the increase of hardness is less than 3%.During corrosion,c-ZrO2?c-ZrO2-x?forms near the interface.A series of crystallographic relationships is detected at ZrC/c-ZrO₂ interface:?11 1^<sub>?_ZrC??1^<sub>1 1^<sub>?_c-ZrO2,[011]_ZrC?[011]_c-ZrO2;?1^<sub>11^<sub>?_ZrC??1^<sub>11^<sub>?_c-ZrO2,?220?_ZrC??220?_c-ZrO2,[011]_ZrC?[011]c-ZrO2.The c-ZrO2 retains some crystallographic orientations of the initial ZrC,presenting an“inheritance in microstructure”between c-ZrO2 and ZrC.The incremental rate of weight gain increased with increased corrosion temperature and decreased C/Zr ratio and the carbon vacancy was passive to the decrease of corrosion rate.The main controlling mechanisms of ZrC ceramics in water vapor changes from phase boundary reactions to surface diffusion control and then to grain boundary diffusion as temperature increased.