Probabilistic Environmentally Assisted Cracking Modeling for Primary Water Stress Corrosion Cracking of Alloy 600

2011 ◽  
pp. 1715-1729
Author(s):  
TaeHyun Lee ◽  
JaeYoung Yoon ◽  
HyoOn Nam ◽  
IlSoon Hwang
Keyword(s):  
Water Stress ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Alloy 600 ◽  
Environmentally Assisted Cracking ◽  
Primary Water

Evaluation and Repair of Primary Water Stress Corrosion Cracking in Alloy 600/182 Control Rod Drive Mechanism Nozzles

2002 ◽  
Author(s):  
Charles R. Frye ◽  
Melvin L. Arey ◽  
Michael R. Robinson ◽  
David E. Whitaker
Keyword(s):  
Water Stress ◽  
Boric Acid ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Base Material ◽  
Corrosion Cracking ◽  
Alloy 600 ◽  
Control Rod ◽  
Drive Mechanism ◽  
Primary Water

In February 2001, a routine visual inspection of the reactor vessel head of Oconee Nuclear Station Unit 3 identified boric acid crystals at nine of sixty-nine locations where control rod drive mechanism housings (CRDM nozzles) penetrate the head. The boric acid deposits resulted from primary coolant leaking from cracks in the nozzle attachment weld and from through-thickness cracks in the nozzle wall. A general overview of the inspection and repair process is presented and results of the metallurgical analysis are discussed in more detail. The analysis confirmed that primary water stress corrosion cracking (PWSCC) is the mechanism of failure of both the Alloy 182 weld filler material and the alloy 600 wrought base material.

Effects of Dissolved Hydrogen on the Primary Water Stress Corrosion Cracking Behavior of Alloy 600 at 325 °C

2004 ◽  
Vol 261-263 ◽  
pp. 943-948 ◽  
Author(s):  
Q.J. Peng ◽  
Tetsuo Shoji
Keyword(s):  
Water Stress ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Experimental Results ◽  
Alloy 600 ◽  
Pressurized Water ◽  
Cracking Behavior ◽  
Primary Water ◽  
Dissolved Hydrogen

Primary water stress corrosion cracking (PWSCC) of Alloy 600 has been a great concern to the nuclear power industry. Reliable PWSCC growth rate data, especially at temperatures in the range of 290-330°C, of the alloy are required in order to evaluate the lifetime of power plant components. In this study, three tests were carried out in simulated pressurized water reactor (PWR) primary water at 325°C at different dissolved hydrogen (DH) concentrations using standard one-inch compact tension (1T-CT) specimens. The initiation and growth of cracks as well as insights into the different PWSCC mechanisms proposed in the literature were discussed. The experimental results show that the detrimental effects of hydrogen on crack initiation and growth reached a maximum at a certain level of DH in water. The experimental results were explained in terms of changes in the stability of the surface oxide films under different DH levels. The experimental results also support the assumption that hydrogen absorption as a result of cathodic reactions within the metal plays a fundamental role in PWSCC.

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