Corrosion Fatigue Behavior Of Nuclear-grade Austenitic Alloys In High Temperature And High Pressure Water

Corrosion fatigue (CF) plays an important role in life design, safety evaluation and life extend in light water reactor (LWR) nuclear power plants (NPPs). Many operation experience and studies indicate CF is one of mainly potential failure issues of environmental assisted cracking for nuclear-grade structural materials. The current ASME code design curve did not fully consider the LWR environmental factors and may have insufficient safety margin or be over-conservative. In 2007, the US Nuclear Regulatory Commission issued RG 1.207, which required a new NPP have to incorporate the environmental effects into fatigue analyses. China is rapidly developing nuclear power and attempting to establish NPPs using domestic technologies. However, there are limited fatigue data of domestic nuclear-grade structural materials in LWR environments, related works are urgent to be done. The present study designs a boat-shaped fatigue specimen to investigate the CF behavior of three types of Alloy 690 steam generator (SG) tubes (tube J, C1 and C2) in borated and lithiated high temperature pressurized water. The CF behavior of domestic forged 316LN stainless steel in high temperature and high pressure water is also investigated, which mainly focus on the effects of strain rate and temperature. The fatigue crack initiation and propagation mechanism of Alloy 690 and 316LN stainless steel in LWR environments are discussed.A kind of boat-shaped fatigue specimen based on thin-wall tubes and a matched fixture were designed and machined. The fixture can clamp the boat-shaped specimen stably. The designed boat-shaped specimen can be employed to evaluate the CF behavior of actual Alloy 690 SG tubes and in high temperature and high pressure water.The CF behavior of actual Alloy 690 SG tubes was investigated in borated and lithiated high temperature pressurized water using the boat-shaped specimens. The fatigue lives of tube J and C1 are comparable in boated and lithiated high temperature pressurized water and locate in lower center region of the ASME code mean curve and above the ASME code design curve, indicating the Tube J and C1 have enough safety margin for fatigue design under present experimental conditions. The fatigue data of the present boat-shaped specimens are longer than those of round-bar specimens published in literatures, which is attributed to the difference of strain ratio (0.2 vs-1) and the difference of size and shape of fatigue specimens. It is believed that the interactions between persistent slip bands (PSBs) and electrochemical factors in high temperature and high pressure water facilitate fatigue crack initiation. The fatigue fracture surface is rough with ridges, on which typical secondary fatigue cracks and fatigue striations have been observed.The CF behavior of three types of Alloy 690 SG tubes was investigated in borated and lithiated high temperature pressurized water using boat-shaped specimens. Main attention was paid to the effects of inclusions on fatigue crack initiation and propagation. The inclusions in tube J, C1 and C2 are mainly TiN and Al2O3/MgO. The inclusion volume fraction for tube J, C1 and C2 are about 0.0894%,0.0384% and 0.1843% respectively. The fatigue life of tube C2 is shorter than the other two due to its highest inclusion volume fraction. The TiN inclusions on surface and TiN-inclusion clusters underneath the surface facilitate fatigue crack initiation. A lot of large TiN inclusions and TiN-inclusion clusters were observed on fracture surface and along main fatigue crack. Furthermore, quasi-cleavage patterns were observed around TiN inclusions. The interactions between large TiN inclusions and hydrogen may accelerate fatigue crack propagation in high temperature and high pressure water.The effects of dissolved oxygen (DO) on CF behavior of Alloy 690 SG tubes were investigated in borated and lithiated high temperature pressurized water. The fatigue life of Alloy 690 is longer at DO=5500 ppb than that at DO<5 ppb. There are more secondary cracks on concave surface for Alloy 690 tubes after fatigue tests in borated and lithiated high temperature pressurized water at<5 ppb DO than that at 5500 ppb DO. The concave surface at<5 ppb DO is rougher than that at 5500 ppb DO. The fatigue crack initiation sites are sunken at<5 ppb DO. The oxide film at<5 ppb DO is much thinner than that at 5500 ppb. Ni and Cr are enriched while Fe is depleted in the oxide film at<5 ppb DO. However, Ni and Fe are enriched while Cr is depleted in the oxide film at 5500 ppb DO. It is believed that the longer fatigue life of Alloy 690 at 5500 ppb DO is attributed to the effects of DO on fatigue crack initiation rather than fatigue crack propagation. The oxide film growth rate is faster at 5500 ppb DO and more Fe ions are released in bulk solution at 5500 ppb DO than at<5 ppb DO, promoting the formation of NiFe2O4. Therefore, the repairation growth rate of broken oxide film is accelerated at 5500 ppb DO, inhibiting selective dissolution and preventing fatigue crack initiation.The fatigue behavior of domestic forged 316LN stainless steel was investigated in air and high temperature and high pressure water. The fatigue life of 316LN stainless steel is shorter in high temperature and high pressure water than that in air. The fatigue data of 316LN stainless steel locate in lower center region of the ASME code mean curve and above the ASME code design curve, indicating it has enough safety margin for fatigue design under the present experimental conditions. The fatigue life of 316LN stainless steel decreases with decreasing strain rate (0.4 %s-1-0.004 %s-1) and keeps almost constant at the strain rate above 0.4 %s-1 in high temperature and high pressure water. The temperature has little influence on fatigue life of 316LN stainless steel in high temperature and high pressure water at the strain rate of 0.4 %s-1. The fracture surface is rough and characterized by multi-crack initiation. The fatigue crack initiation regions appear fan-like patterns. The fatigue crack propagation regions show typical fatigue striations. The fatigue crack mainly initiated at PSBs and rarely at twin boundaries. The dynamic strain aging (DSA) may occur during fatigue tests in LWR environments for 316LN stainless steel. It is believed that the interactions between DSA and LWR environments facilitate fatigue crack initiation and propagation.Based on fatigue data of nuclear-grade austenitic alloys (austenitic stainless steels and nickel-based alloys) in air from our laboratory and open literatures, the Institute of Metal Research (IMR) code mean curves were established using Langer equation. The IMR code design curves were then obtained by reducing the fatigue life at each point on the IMR code mean curve by a factor of 2 on strain or 12 on cycles. Based on fatigue data of nuclear-grade austenitic alloys (austenitic stainless steels and nickel-based alloys) in LWR environments from our laboratory and open literatures, the environmentally assisted fatigue evaluation models were established using environmental fatigue life correction factor (Fen). The Fen mainly considers the effects of strain rate, temperature and DO on fatigue behavior of austenitic alloys in LWR environments. The model can accurately predict the fatigue life of austenitic alloys in LWR environments. An evaluation methodology for environmental fatigue damage of structural materials in NPPs based on environmentally assisted fatigue evaluation models and linear cumulative damage law is proposed.