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

Abstract

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.

Defect sink characteristics of specific grain boundary types in 304 stainless steels under high dose neutron environments

“…At 47.1 dpa, the random HA grain boundaries are at maximum changes in grain boundary concentrations. This trend in the dose dependence of the random HA grain boundaries is consistent with the typically reported segregation trends for neutron irradiated austenitic stainless steels[64].No significant change was observed in FWHM, with increasing dose with the average FWHMbeing 3.9 nm, 4.0 nm, and 3.7 nm for the 5.5, 10.2, and 47.1 dpa samples, respectively, for irradiated random HA grain boundaries. The model predicted FWHMs of 6.8 nm, 7.0 nm and 7.3 nm for the 5.5, Page and 47.1 dpa samples, respectively.…”

“…The random HA grain boundary also shows distinct regions of depletion on either side of the enrichment peak for Ni and a corresponding morphology for Cr. Similar segregation profile morphologies have been observed in neutron and proton irradiated 300 series stainless steels[62][63][64].The experimentally observed RIS response at different grain boundary types was consistent with the predicted RIS response generated from the GiMIK model.Figures 1(b) and 1(c)show the calculated 1D segregation profiles and compare them to the experimentally determined profiles. The model parameters for Fe, Cr, and Ni used in the calculation were taken from Allen et al…”

In situ and tomographic characterization of damage and dislocation processes in irradiated metallic alloys by transmission electron microscopy

“…[1][2][3] It has been shown that the irradiation damage leads to a loss of ductility, an increase in the yield strength and hardness values, and the appearance of a distinct yield point. [4][5][6][7][8] Irradiation can also lead to secondary detrimental effects such as susceptibility to stress corrosion cracking and lower failure strain during creep deformation.…”

Overview of Intergranular Fracture of Neutron Irradiated Austenitic Stainless Steels

“…Another observation of the IG fracture was referred at irradiated Ti stabilized austenitic stainless steel 18Cr-10Ni-Ti, in-service irradiated in WWER reactor core to 2-11 dpa at 260-330 • C. IG fractures were found on fracture mechanics specimens broken to open in nitrogen vapor at about −100 • C (Figures 7 and 8) [3]. Another type of IG fracture appears in IASCC, observed if the irradiated steel is loaded in contact with a high temperature water environment which is used as the primary coolant of BWR/PWR/WWER, e.g., [5,13,25,27,28]. This is a very special type of IG fracture, whose mechanism is not yet sufficiently understood ( Figure 9).…”

“…These results indicated that the intergranular fracture occurrence was dependent on temperature and on specimen stress-strain state. In [17], a compilation of fracture data gathered on A304 and CWA316 irradiated steels was performed [15,17,[20][21][22][23][24][25][26]: it appeared that there are two different conditions for the occurrence of IG modes in stainless steels irradiated in PWRs: Low Temperature High strain Rate (LTHR) condition and High Temperature Low strain Rate (HTLR) condition. The sensitivity to IG mode was higher for higher doses in both conditions, and becomes higher for lower temperature in LTHR conditions and for higher temperature and lower strain rate in HTLR conditions.…”