| Operation experience of nuclear power plants(NPPs)shows that the irradiation assisted stress corrosion cracking(IASCC)of stainless steel(SS)core structural materials has become one major concern for the safe and economical maintenance of NPPs.Previous studies revealed that radiation-induced damages such as dislocation loops,grain boundary segregation and hardening etc.are key material causes for the IASCC.Such damages in the irradiated steel could promote localized deformation adjacent the grain boundaries,which accelerated the crack initiation.To date,our country has realized the localization of core structural materials,while the radiation-induced damage has not been well studied.In addition,the current mechanism of localized deformation promoted IASCC is in need of further improvement.At first,since corrosion is one of the key factors affecting stress corrosion cracking(SCC),a better understanding of the role of corrosion in IASCC is needed while it has not received enough attentions.This requires at first to understand whether and how the irradiation affects the corrosion,and then the corrosion kinetics.Secondly,to fully clarify the effect of localized deformation on IASCC,it further needs to study the synergic effect between the localized deformation and corrosion.In this paper,evolution of radiation-induced damage of domestically fabricated nuclear-grade 304 SS was at first studied,and then the effects of irradiation on corrosion and IASCC of the steel in high temperature water were investigated to clarify the mechanism of IASCC with a focus on the effect of irradiation assisted corrosion on IASCC.The nuclear-grade 304 SS was irradiated to various doses by protons of 2 MeV at 360 °C,followed by characterization of the radiation-induced damage to correlate the evolution of microstructure damage with the irradiation dose.The results revealed that the radiation-induced microstructure damage was mainly dislocation loops with a few micro-voids.The loop density was in the order of 1022 m-3 with an average size of<10 nm.The square root of the product of loop density and size(Nd)0 5,scaled linearly with the square root of irradiation dose with a factor of(6.8 x 103)(dpa)-0.5/mm.The void size was generally less than 5 nm and showed no significant change with the irradiation dose.The loops were believed to be mainly responsible for the hardening in the steel,which also scaled linearly with(Nd)-0.5 with a factor of(1.16 x 10-2)kg/mm.While the depletion of Cr and enrichment of Ni at the dislocation loop and grain boundary showed no difference,the enrichment of Si at the dislocation loop could be of about 6 times of that at the grain boundary.In addition,the loop density and size,as well as radiation-induced segregation(RIS)and hardening were all increased by a higher dose and tended to saturate by a dose of 3-5 displacement per atom(dpa).Effects of irradiation on corrosion of 304 SS in the primary water of pressurized water reactor(PWR)was studied.It was found that increasing the irradiation dose promoted both the overall and intergranular corrosion.Thicknesses of the oxide scale and the inner oxide formed on the surface of 304 SS were both increased by a higher irradiation dose,which was primarily due to the promoted corrosion by irradiation-induced defects.On the other hand,depth of the oxide formed at the grain boundary was also larger for a higher irradiation dose.This was attributed to the RIS,which increased the amount of metal vacancy,the depletion of Cr at the primary grain boundary,and the Ni enrichment adjacent the intergranular oxide.Effects of irradiation on corrosion kinetic of 304 SS in the primary water of PWR was studied.The results revealed that the whole thickness of the oxide scale formed on the steel was increased by a higher exposure time in the short exposure period of<500 h,while a transition of the whole thickness occurred by increasing the exposure period to 1000?1500 h.The was very likely due to the irradiation promoted evolution of the size and density of oxide particles in the outer layer of the oxide scale.The thickness of the inner oxide was increased by a longer exposure time and followed the power law,while the oxidation rate constant increased with increasing the irradiation dose.On the other hand,depth of the intergranular oxide was also larger for a longer exposure time.The intergranular corrosion was related to the evolution of Ni enrichment adjacent the intergranular oxide,which could mitigate the intergranular corrosion by decreasing the diffusion of metal ions at the oxide/metal interface.IASCC behavior of 304 SS in the primary water was studied by interrupted slow strain rate tensile(SSRT)test.The results revealed that the slip step induced by irradiation usually transmitted or terminated at the grain boundary,while the slip step transmitted at grain boundary led to slip continuity at the grain boundary.In contrast,a slip discontinuity was observed at the grain boundary where the slip step terminated,which caused a much higher strain accumulation by feeding dislocations to the grain boundary region.Formation of the slip discontinuity was related to the Schmidt factor pair type of the adjacent grains.Following the SSRT tests,intergranular cracking was observed on surfaces of the irradiated specimens,while the number the cracks were increased by a higher irradiation dose and applied strain,suggesting a higher IASCC susceptibility of the steel in the primary water.Meanwhile,significant intergranular oxidation ahead of the crack tip was observed,while both the width and length of the oxide were larger at a higher irradiation dose.The synergic effect of irradiation-promoted deformation and intergranular corrosion was the primary cause for the IASCC of the irradiated steel.Localized deformation and corrosion in irradiated 304 SS in the primary water of PWR was studied,in an effort to correlate the deformation to corrosion and their synergic effect to cracking.The results revealed that the irradiation promoted the localized deformation at the surface slip step and grain boundary,while the step height as well as the strain concentration at the step and grain boundary was increased by a higher irradiation dose.Increasing the irradiation dose also promoted localized corrosion at the slip step and grain boundary,which was primarily attributed to the strain concentration and lattice distortion induced by irradiation promoted deformation.As the oxidation under stress/strain favored the formation of cracking initiation,a synergic effect of the enhanced localized deformation and corrosion at the slip step and grain boundary caused a higher intergranular susceptibility of the irradiated steel.As stated above,this study firstly investigated the evolution of radiation induced damage of domestically fabricated nuclear-grade SS,which provided the verification data of radiation damage for domestically fabricated nuclear-grade materials.In addition,it quantitatively studied the irradiation assisted corrosion behavior of nuclear-grade SS material,clarified the mechanism of the promotion effect of irradiation on corrosion,and established a quantitative dependence of corrosion behavior of the nuclear-grade SS on irradiation dose.Results of this study aided in clarifying the mechanism of IASCC in terms of the role of the promoted corrosion by irradiation.A data package of the irradiation-induced damage,corrosion and IASCC of domestically fabricated nuclear-grade SS material was obtained,which provided a reliable technical support for optimizing service life of core structural materials and ensuring the safe and economical maintenance of NPPs. |
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