Irradiation Programs and Test Plans to Assess High-Fluence Irradiation Assisted Stress Corrosion Cracking Susceptibility.
. Irradiation assisted stress corrosion cracking (IASCC) is a known issue in current reactors. In a 60 year lifetime, reactor core internals may experience fluence levels up to 15 dpa for boiling water reactors (BWR) and 100+ dpa for pressurized water reactors (PWR). To support a safe operation of our fleet of reactors and maintain their economic viability it is important to be able to predict any evolution of material behaviors as reactors age and therefore fluence accumulated by reactor core component increases. For PWR reactors, the difficulty to predict high fluence behavior comes from the fact that there is not a consensus of the mechanism of IASCC and that little data is available. It is however possible to use the current state of knowledge on the evolution of irradiated microstructure and on the processes that influences IASCC to emit hypotheses. This report identifies several potential changes in microstructure and proposes to identify their potential impact of IASCC. The susceptibility of a component to high fluence IASCC is considered to not only depends on the intrinsic IASCC susceptibility of the component due to radiation effects on the material but to also be related to the evolution of the loading historymore »
- Publication Date:
- OSTI Identifier:
- Report Number(s):
- DOE Contract Number:
- Resource Type:
- Technical Report
- Research Org:
- Idaho National Laboratory (INL), Idaho Falls, ID (United States)
- Sponsoring Org:
- USDOE Office of Nuclear Energy (NE)
- Country of Publication:
- United States
- 36 MATERIALS SCIENCE irradiation assisted stress corrosion cracking
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