Defect Behavior In Several Cladding And Structural Materials:A First-principles Study

Due to the abundant nuclear fuel resources,clean and low energy consumption,nuclear energy(including both fusion and fission)are considered as the most promising solution to human future energy problems.In the harsh environment of a nuclear reactor,plenty of vacancies,self-interstitial atoms,nuclear transmutation products will create in the nuclear materials.The migration and accumulation of those point defects will form voids,dislocation loops and bubbles,leading to swelling and hardening,finally degrading the macroscopic properties of the material and affecting the service life of the material.Therefore,the performance of nuclear materials is a key issue that needs to be solved,which determines the safety of fission reactors and the success of fusion energy commercialization.Although many experiments and simulations have been carried out to investigate the nuclear structural materials,there are still many problems worthy of in-depth study on the fundamental mechanisms of irradiation effect in nuclear materials.Due to excellent high temperature thermal/mechanical properties and good fission product retention capability,zirconium carbide(ZrC)and silicon carbide(3C-SiC)have been considered as important candidate material for the core components of nuclear fuel cladding material in advanced fission reactors,in addition,3C-SiC also deemed as a promising candidate structural material.Oxide dispersion strengthened(ODS)steels and Castable nanostructured alloys(CNAs)are the new generation of RAFM steels,show great promise as the candidate structural materials for nuclear reactors.ODS steel and CNAs containing a high density of fine stable oxide dispersion nanoparticles and MX-type precipitates.The precipitates have a very important effect on the resistance to irradiation damage,and their boundary can capture defects and promote the defect recombination,which can greatly prevent the irradiation swelling and creep in steel.Hence,in this paper,we adopted first-principles calculations to study the behavior of irradiation defects(i.e.vacancy,self-interstitial atom,noble gas atoms,and hydrogen atoms)in these nuclear materials.First,we compared the behavior of intrinsic defects in ZrC?3C-SiC and TaC.The results of intrinsic defect showed that C atoms is easier than Zr atoms escape the lattice site,form vacancy,self-interstitial atom or Frenkel pair in ZrC;C vacancy and anti-site defects are the mean defect in 3C-SiC;form Ta interstitial atom and C atom occupy Ta lattice site cost more energy than other defects in TaC.Second,we investigated the effect of noble gas noble on the crystal lattice.The results indicated that the formation energy of inert gas atoms and the lattice distortion increases linearly with the increase of atomic radius;the binding energy between noble gas atoms are effected by the interstitial configuration and the repulsion of the surrounding lattice atoms;the lattice distortion increases with the number of noble gas atoms,this induces the creation of new vacancies,which in turn trap more noble gas atoms.We compared the behavior of He/H in the different oxide nanoparticles and ?-Fe matrix.The formation energy of single He atom in different oxides is much lower than that of ?-Fe matrix,while single hydrogen shows the opposite phenomenon;He and H diffusion in oxides are more difficult than that in ?-Fe matrix;He atoms tend to occupy different interstitial site individually and H atoms form hydrogen molecule in oxides.The He atoms prefer to occupy adjacent interstitial site and H atoms were more willing to occupy different interstitial site in the MX-type precipitate TaC.The composition and structure of materials significantly affect the behavior of He atoms.In the ODS steel supercell model,He atom is the most stable in oxide Y2O3,followed by the ?-Fe/Y2O3 interface,?-Fe matrix,He atoms prefer to aggregate inside the oxide precipitates rather than larger He bubbles in the ?-Fe matrix;in the CNAs supercell model,He atom is the most stable in ?-Fe/TaC interface,followed by the TaC cluster,?-Fe matrix,the interface as a sink can trap more He atoms;the diffusion of He atom in{100}<110>Fe//{100}<100>TaC interface is easier than that in ?-Fe matrix and TaC;.In this paper,we systematically investigated the behavior of point defects(i.e.vacancy,self-interstitial atom,Frenkel defects,He and H)in the ZrC,3C-SiC,ODS steels,and CNAs,and the effect of the oxidation precipitates,carbide precipitates,grain boundary as well as applied stress on the stability and diffusion of point defect.Our theoretical results not only establish a foundation for understanding the behavior of irradiation defect and the role of precipitate,but also provides some useful input parameters for the next step to study the nucleation and evolution mechanism of gas bubbles in these material with a larger scale simulation method.