Nuclear Executive Update   
An EPRI Progress Report, March 2011
TECHNICAL HIGHLIGHTS
Cooperative Stress Corrosion Cracking Research Program Documents Final Results

The three-phase program, begun in 1995, has advanced the mechanistic understanding of irradiation-assisted stress corrosion cracking in austenitic stainless steels in both boiling and pressurized water reactors.

Irradiation-assisted stress corrosion cracking (IASCC) has affected reactor core internal structures fabricated from austenitic stainless steels in both pressurized water reactors and boiling water reactors. IASCC is influenced by many parameters, including neutron fluence, flux, material, stress or stress intensity factor, temperature, and water chemistry.

The experimental work conducted under the Cooperative IASCC Research Program - a collaborative effort involving utilities, nuclear plant vendors, regulators and research entities from the United States, Europe and Asia - has been compiled in a comprehensive CD (1021235) and final review report (1020986 ). The research complemented other more applied programs by concentrating on developing a better fundamental understanding of physical causes of IASCC, using that understanding to propose better predictive models and identifying potential countermeasures to IASCC.

Key findings include:

  • Modeling IASCC growth rates of irradiated stainless steels in BWR environments will most likely be based on the well known slip/oxidation model. The model should incorporate predictions of the extent of chromium depletion and silicon enrichment at grain boundaries as a function of neutron fluence, as well as the effect of the increase in yield strength and loss of ductility via their effects on crack tip strain rate. An alternative model for IASCC in PWR primary water environments may be necessary.
  • Improved understanding of the mechanistic aspects of IASCC should enable the chemical composition and initial metallurgical state of austenitic stainless steels intended for BWR and PWR internals to be better optimized in the future to resist IASCC. Beneficial characteristics likely include modest levels of cold work that could delay the development of irradiation damage, control of silicon levels to the lower levels of available specifications, presence of grain boundary carbides, and higher nickel alloys that lower the tendency to strain localization.

The Materials Action Plan Committee has developed a roadmap on PWR and BWR Materials Testing to address knowledge gaps on the effects of high fluence on degradation of reactor internals in BWR and PWR environments and to support cost-effective, long-term operation of BWRs and PWRs. EPRI will continue pursuing experimental work through participation in multi-nation irradiated materials testing at the OECD Halden Reactor and collaboration with the U.S. Department of Energy, Idaho National Laboratory and the EDF Materials Aging Institute.

For more information, contact Raj Pathania at 650-855-8762 or rpathani@epri.com.