Stress-corrosion cracking experience in piping of light water reactor power plants

1980 ◽  
Vol 57 (1) ◽  
pp. 133-140 ◽  
Author(s):  
L.C. Shao ◽  
J.J. Burns
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Power Plants ◽  
Reactor Power ◽  
Corrosion Cracking ◽  
Light Water Reactor ◽  
Light Water ◽  
Water Reactor

Probabilistic Fracture Mechanics Benchmarking Study Involving the xLPR and PASCAL-SP Codes: Analysis by PASCAL-SP

2020 ◽  
Author(s):  
Akihiro Mano ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Corrosion Cracking ◽  
Energy Agency ◽  
Pressurized Water ◽  
Water Reactor ◽  
Probabilistic Fracture Mechanics

Abstract A probabilistic fracture mechanics (PFM) analysis code, PASCAL-SP, has been developed by Japan Atomic Energy Agency (JAEA) to evaluate the failure probability of piping within nuclear power plants considering aged-related degradations such as stress corrosion cracking and fatigue for both pressurized water reactor and boiling water reactor environments. To strengthen the applicability of PASCAL-SP, a benchmarking study is being performed with a PFM analysis code, xLPR, which has been developed by U.S.NRC in collaboration with EPRI. In this benchmarking study, deterministic and probabilistic analyses are undertaken on primary water stress corrosion cracking using the common analysis conditions. A deterministic analysis on the weld residual stress distributions is also considered. These analyses are carried out by U.S.NRC and JAEA independently using their own codes. Currently, the deterministic analyses by both xLPR and PASCAL-SP codes have been finished and probabilistic analyses are underway. This paper presents the details of conditions and comparisons of the results between the two aforementioned codes for the deterministic analyses. Both codes were found to provide almost the same results including the values of stress intensity factor. The conditions and results of the probabilistic analysis obtained from PASCAL-SP are also discussed.

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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