PWSCC of Thermocoax Pressurizer Heaters in Austenitic Stainless Steel and Remedial Actions to Preventing SCC

Thébault, Yannick; Moulart, P.; Dubourgnoux, K.; Champredonde, J.; Couvant, Thierry; Neau, Y.; Fageon, J.-M.; Lecharpentier, D.; Breuil, A.; Derouet, V.

doi:10.1007/978-3-319-48760-1_133

Cited by 3 publications

(3 citation statements)

References 0 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…7) in nominal primary water. Pressurizer heaters [38,39], made of stainless steels, and steam generator divider plates [40][41][42], made of rolled Alloy 600 were both affected by SCC. In each case, the level of deformation and heterogeneity at a macroscale, captured by micro hardness or neutron diffraction, was believed to play a major role in the occurrence, location, and kinetics of SCC [43,44].…”

Section: State Of the Artmentioning

confidence: 99%

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Legras¹,

Volgin²,

Radiguet³

et al. 2017

JMSE-B

View full text Add to dashboard Cite

Abstract:Even if temperature, pressure and chemistry of the cooling water are not very high and aggressive, materials used in PWRs (Pressurized Water Reactors) are exposed to different degradation mechanisms. One of the main goals of the research programs in this field is to develop physical model of the mechanisms down to the atomic scale. Such approach needs a clear description and understanding of the degradation mechanisms at the same small scale. This paper illustrates the benefits of different microscopies and of their last improvements up to the promising possibilities of monochromated and aberrations corrected TEM/STEM. A specific focus is placed on four different degradation mechanisms observed in austenitic stainless steel: irradiation ageing, corrosion fatigue, stress corrosion cracking and corrosion.

show abstract

Section: State Of the Artmentioning

confidence: 99%

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Legras¹,

Volgin²,

Radiguet³

et al. 2017

JMSE-B

View full text Add to dashboard Cite

show abstract

“…A significant proportion of the PWR failures under nominally free-flowing conditions have occurred in pressuriser heater sheaths [15,16]. Several factors may make these components more susceptible to intergranular (IG) SCC in non-contaminated primary coolant compared to most other stainless steel components in the primary circuit.…”

Section: Stress Corrosion Crackingmentioning

confidence: 99%

“…These factors include the higher operating temperature compared to the rest of the plant, the use of cold swaging which introduces cold-work and tensile residual stress, thermal cycling introducing dynamic strain and possible fatigue damage which may enhance initiation, heat flux which may cause pH elevation in creviced locations such as at heater support plates or under surface laps, and the presence of surface cold-work. Experimental studies have shown that all the above factors can increase the susceptibility of austenitic stainless steels to IGSCC [15]. Two other failures affecting pressuriser heaters occurred close to the weld connecting the heater to the pressuriser penetration, one at Braidwood Unit 1 in the U.S.A. [17] and one in Sizewell B in the U.K [18].…”

Section: Stress Corrosion Crackingmentioning

confidence: 99%

Contribution of research to the understanding and mitigation of environmentally assisted cracking in structural components of light water reactors

Tice

2017

Corrosion Engineering, Science and Technology

View full text Add to dashboard Cite

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...

show abstract

Stress Corrosion Cracking of Cold Worked Stainless Steels

Cattant

2022

Materials Ageing in Light-Water Reactors

View full text Add to dashboard Cite

PWSCC of Thermocoax Pressurizer Heaters in Austenitic Stainless Steel and Remedial Actions to Preventing SCC

Cited by 3 publications

References 0 publications

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Contribution of research to the understanding and mitigation of environmentally assisted cracking in structural components of light water reactors

Stress Corrosion Cracking of Cold Worked Stainless Steels

Contact Info

Product

Resources

About