Crack Growth Behavior of Irradiated Type 316 SS in Low Dissolved Oxygen Environment

Chen, Y.; Alexandreanu, B.; Yang, Yong; Shack, W. J.; Natesan, K.; Gruber, E.E.; Rao, A. S.

doi:10.1002/9781118456835.ch121

Cited by 1 publication

(1 citation statement)

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“…The saturation dose for CGR seemed to be lower than that for initiation stress in BWR water conditions as shown in Figure 14. The dependence of CGR on stress intensity K is expressed as the nth power of K. The n values in the literature were within 1.5-2.5 in BWR water conditions [7], while some data showed lower n values of 1.2 at higher doses in MTR data [200,203]. These data suggested that K dependence might be altered by irradiation and became smaller at higher dose.…”

Section: Crack Growthmentioning

confidence: 93%

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

Fukuya¹

2013

Journal of Nuclear Science and Technology

153

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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.

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Section: Crack Growthmentioning

confidence: 93%