Review of PWSCC and mitigation management strategies of Alloy 600 materials of PWRs

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 le...

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Bibliographic Details
Published in:Journal of nuclear materials 2013-11, Vol.443 (1-3), p.321-330
Main Author: Hwang, Seong Sik
Format: Article
Language:English
Subjects:
Applied sciences
Controled nuclear fusion plants
Energy
Energy. Thermal use of fuels
Exact sciences and technology
Fission nuclear power plants
Fuels
Heaters
Inspection
Installations for energy generation and conversion: thermal and electrical energy
Maintenance
Management
Nozzles
Nuclear fuels
Penetration
Pressurized water reactors
Strategy
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Summary: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.
ISSN:0022-3115
1873-4820
DOI:10.1016/j.jnucmat.2013.07.032