“…The sensitization (Cr depletion of grain boundaries) results in intergranular fracture of austenite steels only in oxygen containing water envi ronments, whereas this has no effect on IGSCC in an oxygen free water environment [28,29]. Hardening and not sensitization is the requirement for initiation of susceptibility to IGSCC in an oxygen free water environment.…”
A physicomechanical model for occurrence of fractures of austenitic steels upon intergranular stress corrosion cracking (IGSCC) is proposed on the basis of an analysis of experimental data on the suscep tibility of irradiated austenitic steels to IGSCC during contact with the coolant of the primary circuit of a water moderated water cooled power reactor (VVER). A durability evaluation method for units of vessel internals by the criterion of the initialization of IGSCC is developed that allows for calculations under a time varying stress strain state of the unit and a damaging radiation dose.
“…The sensitization (Cr depletion of grain boundaries) results in intergranular fracture of austenite steels only in oxygen containing water envi ronments, whereas this has no effect on IGSCC in an oxygen free water environment [28,29]. Hardening and not sensitization is the requirement for initiation of susceptibility to IGSCC in an oxygen free water environment.…”
A physicomechanical model for occurrence of fractures of austenitic steels upon intergranular stress corrosion cracking (IGSCC) is proposed on the basis of an analysis of experimental data on the suscep tibility of irradiated austenitic steels to IGSCC during contact with the coolant of the primary circuit of a water moderated water cooled power reactor (VVER). A durability evaluation method for units of vessel internals by the criterion of the initialization of IGSCC is developed that allows for calculations under a time varying stress strain state of the unit and a damaging radiation dose.
“…A number of tests have clearly shown that the presence of sulphur near the upper specification limit for Type 304 or 304L stainless steel severely retards crack growth, as shown in Figure 10 [20]. More recent tests have shown that even sulphur contents in the range 0.004-0.006% have an appreciable effect in retarding crack growth rate, as shown in Figure 11 [22]. Figure 10 Crack growth rates for steels with different sulphur content under constant load.…”
Section: Stress Corrosion Crackingmentioning
confidence: 99%
“…Lower temperature components are, thus, less affected, as are those made from material given a thermal treatment at around 700°C which improves the microstructure by precipitating carbides on grain boundaries. Nickel-based weld metals, such as Alloys 182 and 82, are also susceptible to PWSCC [32,33], as are higher strength precipitation-hardened alloys such as Alloy-X-750 [22]. A variety of mechanisms can cause secondary side cracking including caustic environments or acidic conditions, both of which may occur in crevice locations under certain conditions [34].…”
Environmentally assisted cracking (EAC) is a potential threat to the safety and integrity of water-wetted components in operating water-cooled nuclear power plants. Two forms of EAC are commonly distinguished, depending on the form of loading contributing to damage: stress corrosion cracking (SCC) and corrosion fatigue. A number of instances of in-service degradation due to EAC have occurred in operating plants worldwide, often leading to unplanned plant outages. Understanding the causes of EAC is essential to minimise the loss of plant availability due to its occurrence and to avoid the possibility of catastrophic failure, for example, if a crack grew to a critical size in a major pressure boundary component. This paper will describe some examples of these phenomena in the main materials of construction of pressure boundary and other critical components in pressurised and boiling water reactors (BWRs). Over the last several decades, substantial research programmes have been carried out in a number of laboratories worldwide, aimed at further understanding of the processes leading to EAC to manage occurrences in plant and minimise future failures. Selected areas of research on EAC in light water reactor environments are discussed. Corrosion fatigue in low-alloy pressure vessel steels was the subject of considerable attention in the 1980s and early 1990s because of its potential threat to pressure vessel integrity and the publication of data, suggesting that there is a major influence of environment on fatigue crack growth in some laboratory tests. The author's research provided insight into the conditions under which the major environmental effects occur and contributed to the development of an ASME Code Case for pressurised water reactor (PWR) conditions which provided a means of screening based on steel sulphur content and loading conditions. More recently, the research focus in this area has moved to austenitic stainless steels, again providing support to Code Case development and furthering mechanistic understanding. A recent review of knowledge gaps for EPRI provides a basis for future research on environmentally assisted fatigue and will inform the development of new assessment methodologies. A key area of the current study concerns differences in loading conditions between specimens in laboratory tests and plant components subject to transient loading. In the case of SCC, stainless steels have shown the greatest propensity to cracking in BWRs, while Alloy 600 has been a major cause of in-service failures in PWRs, both on the primary side, as recognised by Coriou in the early 1960s, and in secondary environments where a number of different corrosionrelated failure processes have been identified. High-strength alloys, such as Alloy X-750 used for fastener applications, have also caused failures in both reactor types. For austenitic materials, SCC susceptibility is enhanced by irradiation, resulting in failures in core internals components. Ferritic stainless steels also undergo SCC under some specific circums...
“…This condition renders it "sensitized," susceptible to Intergranular Stress Corrosion Cracking under tensile stress in an environment promoting corrosion [7]. The barrier hardening in the matrix caused during the irradiation likely causes additional strain at the grain boundaries during deformation that is less likely to be accommodated in the sensitized microstructure.…”
Section: Mechanical Properties Of Low Temperature Irradiated Stainlesmentioning
Discovery of aging phenomena in the materials of a structure may arise after its design and construction that impact its structural integrity. This condition can be addressed through a demonstration of integrity with the material-specific degraded conditions. Two case studies of development of fracture and crack growth property data, and their application in development of in-service inspection programs for nuclear structures in the defense complex are presented. The first case study covers the development of fracture toughness properties in the form of J-R curves for rolled plate Type 304 stainless steel with Type 308 stainless steel filler in the application to demonstrate the integrity of the reactor tanks of the heavy water production reactors at the Savannah River Site. The fracture properties for the base, weld, and heat-affected zone of the weldments irradiated at low temperatures (110°-150°C) up to 6.4 dpa NRT and 275 appm helium were developed. An expert group provided consensus for application of the irradiated properties for material input to acceptance criteria for ultrasonic examination of the reactor tanks. Dr. Spencer H. Bush played a lead advisory role in this work. The second case study covers the development of fracture toughness for A285 carbon steel in high level radioactive waste tanks. The approach in this case study incorporated a statistical experimental design for material testing to address metallurgical factors important to fracture toughness. Tolerance intervals were constructed to identify the lower bound fracture toughness for material input to flaw disposition through acceptance by analysis.
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