In-reactor stress relaxation in bending of 20% cold-worked 316 stainless steel

Kenfield, T.A.; Busboom, H.J.; Appleby, W.K.

doi:10.1016/0022-3115(77)90061-7

Cited by 12 publications

(1 citation statement)

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“…Understanding on irradiation creep has obtained mainly from data at temperatures higher than 400 • C [278] and experimental data at LWR-relevant temperatures (280-370 • C) are scarce [279,280]. Recently data on irradiation creep and stress relaxation at temperatures around 300 • C have been accumulated by MTR irradiation experiments in the last decade [281][282][283][284][285][286][287].…”

Section: Irradiation Creep and Stress Relaxationmentioning

confidence: 99%

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

Fukuya¹

2013

Journal of Nuclear Science and Technology

161

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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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