Review of PWSCC and mitigation management strategies of Alloy 600 materials of PWRs
Source abstract
Abstract
Abstract Primary water stress corrosion cracking (PWSCC) of Alloy 600 penetration nozzles in pressurized water reactors (PWRs) was reported in the control rod drive mechanism (CRDM), pressurizer instrumentation, and pressurizer heater sleeves. Recently, two cases of boric acid precipitation that indicated leaking of the primary cooling water were reported on the bottom head surface of the steam generators (SGs) in Korea. PWSCC crack indication in CRDM was also detected in a Korea plant. It is necessary to set up a rigid maintenance and inspection guidelines for the components. The PWSCC history of Alloy 600 penetration nozzles of PWRs and maintenance strategies are reviewed based upon the open literature and some experiences in Korea. The inspection requirements, repair techniques such as material changes, the isolation, weld overlays, stress improvements, water chemistry changes are reviewed. Management strategies for the Alloy 600 nozzles are also described.
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Research context
OpenAlex topics: Nuclear Materials and Properties · Non-Destructive Testing Techniques · Hydrogen embrittlement and corrosion behaviors in metals
SciVX discovery tags are terminology-based. OpenAlex topics are source-provided classifications.
References (47)
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