Evolution of fine-scale defects in stainless steels neutron-irradiated at 275 C

Current phenomenological knowledge and understanding of mechanisms are reviewed for radiation embrittlement of reactor pressure vessel low alloy steels and irradiation assisted stress corrosion cracking of core internals of stainless steels. Accumulated test data of irradiated materials in light water reactors and microscopic analyses by using state-of-the-art techniques such as a three-dimensional atom probe and electron backscatter diffraction have significantly increased knowledge about microstructural features. Characteristics of solute clusters and deformation microstructures and their contributions to macroscopic material property changes have been clarified to a large extent, which provide keys to understand in the degradation mechanisms. However, there are still fundamental research issues that merit study for long-term operation of reactors that requires reliable quantitative prediction of radiation-induced degradation of component materials in low-dose rate high-dose conditions.

https://www.tandfonline.com/doi/pdf/10.1080/00223131.2013.772448

Current understanding of radiation-induced degradation in light water reactor structural materials

Isolating the effect of radiation-induced segregation in irradiation-assisted stress corrosion cracking of austenitic stainless steels

Deformation mode map of irradiated 316 stainless steel in true stress–dose space

Irradiation-assisted stress corrosion cracking (IASCC) is aptly named since the effect of irradiation is to enhance an inherent susceptibility to stress corrosion cracking (SCC). This chapter introduces the basic SCC dependencies in austenitic stainless steels and nickel alloys under unirradiated conditions and then describes how they are accentuated or diminished as a result of radiation. SCC results from a confluence of stress, microstructure, and water chemistry, and each is affected by irradiation. With increasing plant operation and improved laboratory capability, it has been concluded that true immunity to SCC growth apparently does not exist in common engineering materials although different conditions can produce large changes in SCC susceptibility. As nuclear power plants operate longer, an increased incidence of SCC can be expected unless active mitigation steps are taken.

F Gröschel

Papers

The effects of irradiation and testing temperature on tensile behaviour of stainless steels