Nuclear Engineering Division

Corrosion and Mechanics of Materials

Light Water Reactors

 

To continue safe operation of current LWRs, the aging degradation of the reactor structures must be adequately understood and managed. Potential aging mechanisms include fatigue and environmentally assisted cracking of piping and pressure vessels, and irradiation-assisted stress corrosion cracking (IASCC) of reactor internals. Nonsensitized austenitic stainless steels (SSs) become susceptible to intergranular failure after accumulation of a sufficient neutron fluence. Such cracking has occurred in control-blade sheaths and handles and in instrument dry tubes of boiling water reactors (BWRs). Intergranular cracking has also occurred in more safety-significant structural core components in BWRs, such as the top guide, shroud, and core plate. However, the relative contributions of neutron fluence, material composition, heat-treatment condition (sensitization), and fabrication variables (welding method and residual weld and fit-up stresses) to crack initiation and growth are not clear. The current LWR research is focused on:

  1. fatigue of pressure vessel and piping steels,
  2. crack growth in austenitic SSs,
  3. IASCC of austenitic SSs, and
  4. environmentally assisted cracking in high-nickel alloys.

Corrosion and Mechanics of Materials: Light Water Reactors
Next page: Fatigue Testing of Carbon Steels and Low–Alloy Steels
1 2  3 4 5 6 Next »

Last Modified: Thu, April 21, 2016 4:55 AM

 

RELATED RESOURCES

Postdoc Jobs

For more information:

Corrosion and Mechanics of Materials
Sect. Manager: K. Natesan

K. Natesan's Executive Bio