Stress Corrosion Cracking Behavior Of Alloy 690 And 600 In High Temperature Water

Stress corrosion cracking(SCC)of the structural materials has always been a threat to the safety and economy of nuclear power plants(NPPs)because of the severe damage it has caused to the reactor coolant pressure boundaries during long-term operation in the corrosive high temperature water environment.Starting in the late 1960 s,Generation II reactors suffered major economic losses from large scale failures of alloy 600 steam generator tubes.New advanced reactors(from Generation II+)replaced the alloy 600 with alloy 690.Extensive work has focused on the SCC behavior of alloy 600 and its replacement,alloy 690,to identify the SCC dependencies and mechanisms,and to establish a service-life prediction model.Although qualitative and semiquantitative data obtained from U-bends,C-rings and slow strain rate tensile(SSRT)tests initially contributed to the material selection and water chemistry control for NPPs,more accurate,reliable and quantitative experimental data for crack initiation time,cracking thresholds and crack growth rates(CGRs)are urgently needed for in-service aging management and life prediction for NPPs' long-term operation.So far,these accurate and reliable quantitative SCC data are rare due to the many technical complexities and the high cost of SCC experiments in high-temperature and high-pressure water.The poor understanding of the contributing factors(material condition,water chemistry and load)and underlying SCC mechanisms remains,and the lack of acceptable prediction models continues to be a major challenge.In this study,the SCC initiation and propagation behavior of alloy 690 and 600 in simulated reactor coolant water was investigated.First,a comparative study of crack initiation and propagation behavior of these two alloys with different heat treatments was conducted to obtain quantitative experimental data on crack initiation time and CGR,which provide solid support for the proposed prediction model.Then,the effects of cold work,grain boundary carbides,dissolved oxygen(DO),dissolved hydrogen(DH),impurity ions(Cl-)and crack tip stress intensity factor(K)on CGR were evaluated to account for key combinations of material condition,water chemistry and loading.Finally,the SCC mechanisms of alloy 600 and 690 were elucidated using highresolution microstructural characterization techniques.A corresponding CGR prediction model for alloy 690 was also established.The major conclusions are as follows:(1)DCPD on-line monitoring provided high resolution data on the evolution of cracking,and thus permitted very accurate determination of the time to SCC initiation.A good correlation between crack initiation time and CGR data was identified,and similar mechanisms for SCC initiation and propagation were revealed.From these data and insights,an empirical model was developed to predict the crack initiation time for cold-worked alloy 600 in pressurized water reactor(PWR)primary water.(2)The relationship between grain boundary(GB)carbides and cold work on SCC behavior was elucidated.The existence of carbides is beneficial for non-cold worked alloy 600,which is attributed to reduced GB sliding and GB internal oxidation.However,for cold worked alloy 600,GB carbides both increased GB local strain and accelerated GB internal oxidation,which is detrimental.For alloy 690,cold work and GB carbides accelerated CGR due to increased crack tip strain rate,either from increased yield strength or higher residual strain at GBs.(3)The effect of DO and DH on SCC behavior was evaluated.The CGR of both alloys in hydrogenated water(at the Ni/Ni O phase boundary)was higher than that in oxygenated water.For alloy 600,the effect of DO and DH was strongly related to the extent of GB internal oxidation.The faster CGR in hydrogenated water(up to 200 times higher)can be attributed to more severe GB oxidation.However,for alloy 690,the GB internal oxidation model is not able to explain the SCC behavior because no GB oxidation was observed.On the other hand,a good correlation between the GB cavity coverage and high CGRs was observed in cold worked alloy 690 when tested in hydrogenated water.This indicates that DH might promote cavity formation ahead of the crack tip,and thus decrease the GB strength and cause an enhancement in the CGR up to 4 times.(4)Effects of loading under ±d K/da and ±d K/dt conditions on CGR were quantified,and the exponential relationship between CGR and K was obtained by fitting the CGR-K curves.Positive d K/da accelerates the CGR up to 362 times when compared with constant K and positive d K/dt,which was attributed to the higher crack tip strain rate.The accelerating effect of +d K/da on CGR was underestimated by existing SCC models.Dropping K by-d K/dt might produce misleading and non-conservative data because cracks in essentially all components grow under d K/da conditions.Thus,d K/da is strongly recommended for measuring of KISCC.(5)For alloy 690,a mechanical(stress)dominated SCC mechanism in PWR primary water was clarified.The effect of mechanical properties changes caused by cold work and GB carbides on CGR was much greater than the effect of environmental factors(corrosion potential,DO,DH,impurity ions,etc.).Finally,a SCC CGR prediction model for cold worked alloy 690 was proposed that addresses the effects of temperature,cold work,DH,DO,GB carbides and stress intensity factor K.