Investigations Of Ion Irradiation-induced Damage Evolution In Pre-hydrided Zr-based Alloys

Zirconium alloys are widely used as cladding materials in pressurized water reactors due to their low neutron absorption cross-section,excellent corrosion resistance and mechanical properties at operating temperature.The power plants are using longer cycles to push the burn-ups to higher levels,leading to a higher neutron fluences and higher hydrogen concentration of zirconium claddings during corrosion reaction with the primary water environment.Therefore,the investigations of ion irradiation-induced damage evolution in pre-hydrided Zr-based alloys with more comprehensive condideration of in-plie environment would provide a better understanding of irradiation damage mechanism and contribute to the performance evaluation and lifetime prediction of zirconium alloys,which is significant in nuclear industry.The materials investigated in the study is pre-hydrided Zircaloy-4 with hydrogen content of 500 ppm,which was irradiated with 4.6 MeV Kr17+ at room temperature and high temperature.The ion irradiation-induced damage evolution in pre-hydrided Zircaloy-4 was investigated with scanning electron microscopy,atomic force microscopy,transmission electron microscopy and slow positron annihilation technique,including:morphology evolution under hydrogen desorption process and irradiation,irradiation induced microstructure evolution at room temperature and high temperature.Firstly,the morphology evolution under hydrogen desorption and continuous irradiation was investigated to establish the microstructure evolution in hydrogen desorption process and to propose the mechanism of surface hydride convex,the irradiation-induced surface roughening mechanism and irradiation-induced hydrogen release acceleration mechanism combined with morphology evolution of α-Zr matrix and surface hydride.The surface hydride convex forms in the hydrogen desorption process,which consists of α-Zr fine grains in the surface region and a significant increase in the amount of δ-hydrides is observed in the sub-surface region.The hydrogen diffuses from the internal to the surface,leading to the hydrogen enrichment in the sub-surface region.In the subsequent cooling process,δ-hydride precipitates with volume expansion of 17.2%per unit cell in the phase transformation of α-Zr to δ-hydride,which should lead to the formation of convex.However,the ion irradiation restrains the formation of surface hydride convex based on the irradiation-induced hydrogen release acceleration mechanism.On the one hand,the irradiation damages would create diffusion channel for soluble hydrogen atoms to release from the surface.On the other hand,the displacement of Zr atoms would result in the break of Zr-H bonds.The increasing Ra value indicates the gradural surface roughening process with continuous irradiation and the irradiation induced surface damage is primarily in the form of damaged pits.The irradiation induced surface roughening mechanism is suggested as three stages:stage I,the formation and growth along depth direction of damaged pits at low doses;stage II,the extension of damaged region within the surface;stage III,layer by layer removal of surface atoms once the undamaged matrix is depleted.Then,the microstructure evolution at room temperature is studied,including dislocation loops,second-phase particles,hydride and Kr bubbles.The results showed that,the distribution of a-loops and c-loops was small-sized but high density in α-Zr matrix.The complete amorphization of SPPs occurs at the early stage of ion irradiation with the redistribution of O element from SPPs into α-Zr.The nano-sized hydride phases were formed in α-Zr matrix due to the hydrogen trapping and aggregation in irradiation-induced defects.The density of Kr bubbles shows a positive correlation with Kr concentration,however,the size is irrelevant with Kr concentration.Finally,the microstructure evolution at high temperature is studied,including the evolution of dislocation loops,second-phase particles,hydride,Kr bubbles with irradiation doses and element segregation in grain boundary.The corduroy stripes were observed in the surrounding of partically-dissolved SPPs after irradiation dose of 20 dpa.The nature of corduroy stripes was irradiation defects with c-loop features.The SPPs undergo irradiation-induced dissolution process,of which the density and average size decreases,meanwhile,accompanied with Fe,Cr,O depletion.The hydride phase was subjected to dissolve firstly and then precipitate after irradiation.The ζ-hydride as transient phase precipitates in α-δphase boundary to reduce the lattice strain.Significant Fe and Sn segregation in the irradiated grain boundaries was observed,which is the evidence of irradiation-induced segregation.