General corrosion and stress corrosion cracking of Alloy 600 in light water reactor primary coolants

2019 ◽  
Vol 524 ◽  
pp. 340-375 ◽  
Author(s):  
Peter M. Scott ◽  
Pierre Combrade
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Light Water Reactor ◽  
General Corrosion ◽  
Alloy 600 ◽  
Light Water ◽  
Water Reactor

scholarly journals Resistance of Ferritic FeCrAl Alloys to Stress Corrosion Cracking for Light Water Reactor Fuel Cladding Applications

CORROSION ◽  
10.5006/3632 ◽  
2020 ◽  
Vol 76 (11) ◽  
Author(s):  
Raul B. Rebak ◽  
Liang Yin ◽  
Peter L. Andresen
Keyword(s):  
High Temperature ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Fuel Cycle ◽  
Austenitic Stainless Steels ◽  
Corrosion Cracking ◽  
Fuel Cladding ◽  
Light Water ◽  
Water Reactor ◽  
Fecral Alloys

Since 2011, the international nuclear materials community has been engaged in finding replacements for zirconium alloys fuel cladding for light water reactors. Iron-chromium-aluminum (FeCrAl) alloys are cladding candidates because they have high strength at high temperature and an extraordinary resistance to attack by superheated steam in the event of a loss of coolant accident. As FeCrAl alloys have never been used in nuclear reactors, it is important to characterize their behavior in the entire fuel cycle. Stress corrosion cracking (SCC) studies were conducted for two FeCrAl alloys (APMT and C26M) in typical simulated boiling water reactor conditions at 288°C containing either dissolved hydrogen or oxygen. Crack propagation studies showed that both ferritic FeCrAl alloys were resistant to SCC at stress intensities below 40 MPa√m. Current results for FeCrAl confirm previous findings for Fe-Cr alloys showing that ferritic stainless alloys are generally much more resistant to high-temperature water SCC than austenitic stainless steels.

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