Jibo Tan, Xinqiang Wu, En-Hou Han
CAS Key Laboratory of Nuclear Materials and Safety Assessment, Institute of Metal Research,
Chinese Academy of Sciences, Shenyang, 110016, PR China 

EXTENDED ABSTRACT: The nuclear energy development route is hot reactors--+fast reactors--+fusion reactors. Pressurized water reactor (PWR) is the main type of reactor at present, and the fourth-generation advanced reactors are also developing rapidly, such as the lead-cooled fast reactors (LFR). The typical service environment of PWR is high­temperature pressurized water, and of LFR is high-temperature liquid lead bismuth eutectic (LBE). Due to the shutdown/ start-up, thermal shock and thermal delamination, key equipment materials suffer low cycle fatigue damage during service. Meanwhile, high-temperature pressurized water corrosion, liquid metal corrosion, liquid metal embrittlement (LME) interact with the alternating load induce corrosion fatigue damage. ASME code mainly gives the fatigue design curve of key equipment materials (austenitic stainless steels, T91 ferritic-martensitic steel, etc.) in air, without fully considering the effects of enviromnental factors. In this paper, a series of in-situ measurement technologies were developed, such as strain and crack length measurement technology in high-temperature pressurized water and liquid LBE, and corrosion fatigue testing equipments were developed. The low cycle fatigue behavior of austenitic stainless steels (316LN, 304L, 308L and their welds) in high-temperature pressurized water was studied. The influence of material types, surface finish, temperature, strain rate, oxygen concentration and other factors on their corrosion fatigue life was obtained, and the enviromnental fatigue life predicted model was established. The corrosion fatigue behavior of typical candidate materials (T9 l steel and 3 l 6LN stainless steel) in LFR was studied. The liquid metal corrosion and embrittlement related to temperature, strain amplitude and strain rate, etc., are the key factors affecting the corrosion fatigue life of materials. The fatigue properties of T9 l steel significantly degraded due to LME, while 3 l 6LN stainless steel is less sensitive to LME and its fatigue life is only slightly reduced at high strain amplitude. The LME mechanism of T91 steel during fatigue is Pb/Bi permeation and segregation induced superstructure at sub-grain boundary. Enviromnental fatigue life predicted models of structural materials in liquid LBE are being developed.
Keywords: Nuclear materials; Corrosion fatigue; High-temperature pressurized water; liquid heavy metal
REFERENCES
[l] Min Wei, Lei Zhao etc., Acta Mater., 261, (2023) 119365