Austenitic stainless steels are used extensively as structural materials in the internal components of reactor pressure vessels. However, high neutron doses lead to a significant reduction in the fracture resistance of these steels in water environment. Irradiation assisted stress corrosion cracking (IASCC) of internals has been observed in pressurized water reactors (PWRs).
In the present work the IASCC model of the irradiated austenitic steels in PWR water has been developed.
On the basis of analysis of available experimental data IASCC mechanism is proposed. Based on this mechanism, the dependence of fracture stress under IASCC on neutron dose is derived. For its construction the following assumptions were made.
1. Creep rate due to grain boundary sliding does not depend on neutron dose.
2. Fracture strain due to grain boundary sliding decreases when neutron dose increases.
3. There is an apparent stress threshold below which IASCC initiation does not occur in PWR environment.
Life prediction analysis for IASCC is performed on the basis of linear rule of damage accumulation.