Abstract:During normal operation light water reactor (LWR) pressure vessels are subjected to a variety of transients resulting in time varying stresses. Consequently, fatigue and environmentally assisted fatigue are growth mechanisms relevant to flaws in these pressure vessels. In order to provide a better understanding of the resistance of nuclear pressure vessel steels to flaw growth process, a series of fracture mechanics experiments were conducted to generate data on the rate of cyclic crack growth in SA508-2 and S… Show more
“…Numerous studies had shown that steel sulphur content was an important factor influencing environmental enhancement of fatigue crack growth in PWR environments, e.g. Figure 18 [55]. Different crack growth curves are, therefore, provided in the Code Case depending on the sulphur content of the steel: no enhancement for <0.004%S, intermediate enhancement for 0.004%≤S ≤ 0.013%, and the largest degree of enhancement for S > 0.013%.…”
Section: Corrosion Fatiguementioning
confidence: 99%
“…The enhanced crack growth curves are required to be used only if both the incremental time-based crack growth rate (d a /d t ) and crack growth increment (Δ a ) exceed critical values. Figure 18 Summary of corrosion fatigue crack growth data for low-alloy steels in PWR and BWR environments showing that steel sulphur content has a much larger effect under PWR conditions [55]. …”
Section: Corrosion Fatiguementioning
confidence: 99%
“…Summary of corrosion fatigue crack growth data for low-alloy steels in PWR and BWR environments showing that steel sulphur content has a much larger effect under PWR conditions [55]. …”
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...
“…Numerous studies had shown that steel sulphur content was an important factor influencing environmental enhancement of fatigue crack growth in PWR environments, e.g. Figure 18 [55]. Different crack growth curves are, therefore, provided in the Code Case depending on the sulphur content of the steel: no enhancement for <0.004%S, intermediate enhancement for 0.004%≤S ≤ 0.013%, and the largest degree of enhancement for S > 0.013%.…”
Section: Corrosion Fatiguementioning
confidence: 99%
“…The enhanced crack growth curves are required to be used only if both the incremental time-based crack growth rate (d a /d t ) and crack growth increment (Δ a ) exceed critical values. Figure 18 Summary of corrosion fatigue crack growth data for low-alloy steels in PWR and BWR environments showing that steel sulphur content has a much larger effect under PWR conditions [55]. …”
Section: Corrosion Fatiguementioning
confidence: 99%
“…Summary of corrosion fatigue crack growth data for low-alloy steels in PWR and BWR environments showing that steel sulphur content has a much larger effect under PWR conditions [55]. …”
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...
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