Mechanical engineering principles

Paine, Robert T.; Beards, Christopher; Tucker, Pamela D.; Bacon, D.H.

doi:10.1016/b978-0-08-052398-9.50005-6

Cited by 2 publications

(1 citation statement)

References 0 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The activation energy for crack initiation of Alloy 600 was suggested by others. [20][21][22][23][24][25] For a susceptibility prediction in this study, the value of 210 KJ/mole was selected for crack initiation of Alloy 600 with reasonable consensus from six studies. 20-25…”

Section: Comparison With Alloy 690 Nozzle At Hot Leg Sidementioning

confidence: 99%

Primary Water Stress Corrosion Cracking Analysis in Alloy 600 Steam Generator Nozzle of a Pressurized Water Reactor

Hwang

Lim²,

Kim³

2013

Corrosion

View full text Add to dashboard Cite

Steam generator (SG) drain nozzles on the cold leg side in a pressurized water reactor (PWR) have exhibited stress corrosion cracking (SCC). The nozzles were made of forged and annealed Alloy 600 (UNS N06600) base material containing welds with Alloys 82 (UNS N06082) and 182 (UNS W86182). Carbides were distributed at the grain boundaries, but no grain boundary Cr depletion was observed in the base metal. Circumferential cracks developed at the root of a J weld inside a thick wall pipe. Axial cracks were distributed with a length around 6 mm to 10 mm long. All cracks were initiated from the inner surface of the pipe. Some more shallow cracks were also observed, but they were not detected by the field non-destructive examination using eddy current test (ECT) and ultrasound test (UT). Two out of the 12 cracks penetrated through the pipe wall, and therefore the primary water was leaking during the operation. The high tensile stress region calculated by a finite element modeling (FEM) model coincided with the crack locations. The highest tensile stress region was at the root of the J weld, as judged from metallography and residual stress analyses. Residual tensile stresses from welding are considered to be the main contributor to the overall stress.

show abstract

Section: Comparison With Alloy 690 Nozzle At Hot Leg Sidementioning

confidence: 99%

Primary Water Stress Corrosion Cracking Analysis in Alloy 600 Steam Generator Nozzle of a Pressurized Water Reactor

Hwang

Lim²,

Kim³

2013

Corrosion

View full text Add to dashboard Cite

show abstract

Corrosion control of nuclear steam generators under normal operation and plant-outage conditions: a review

Rodríguez

2020

Corrosion Reviews

View full text Add to dashboard Cite

AbstractNuclear steam generators (NSGs) are key components of nuclear power plants (NPPs) since their reliability affects the overall plant performance. Corrosion control of steam generators under operation and during plant outages is attained by proper design, adequate material selection and control of water chemistry in the entire secondary cooling circuit. This article reviews the types of steam generators, the materials used for their construction (with a special focus on tubing alloys), the corrosion processes that may occur under operation and the strategies to preserve steam generators and other components of the secondary cooling system from corrosion during plant outages.

show abstract

Mechanical engineering principles

Cited by 2 publications

References 0 publications

Primary Water Stress Corrosion Cracking Analysis in Alloy 600 Steam Generator Nozzle of a Pressurized Water Reactor

Primary Water Stress Corrosion Cracking Analysis in Alloy 600 Steam Generator Nozzle of a Pressurized Water Reactor

Corrosion control of nuclear steam generators under normal operation and plant-outage conditions: a review

Contact Info

Product

Resources

About