Crack growth rates and fracture toughness of irradiated austenitic stainless steels in BWR environments.

[摘要] In light water reactors, austenitic stainless steels (SSs) are used extensively as structural alloys in reactor core internal components because of their high strength, ductility, and fracture toughness. However, exposure to high levels of neutron irradiation for extended periods degrades the fracture properties of these steels by changing the material microstructure (e.g., radiation hardening) and microchemistry (e.g., radiation-induced segregation). Experimental data are presented on the fracture toughness and crack growth rates (CGRs) of wrought and cast austenitic SSs, including weld heat-affected-zone materials, that were irradiated to fluence levels as high as {approx} 2x 10{sup 21} n/cm{sup 2} (E > 1 MeV) ({approx} 3 dpa) in a light water reactor at 288-300 C. The results are compared with the data available in the literature. The effects of material composition, irradiation dose, and water chemistry on CGRs under cyclic and stress corrosion cracking conditions were determined. A superposition model was used to represent the cyclic CGRs of austenitic SSs. The effects of neutron irradiation on the fracture toughness of these steels, as well as the effects of material and irradiation conditions and test temperature, have been evaluated. A fracture toughness trend curve that bounds the existing data has been defined. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components have also been evaluated.

[发布日期] 2008-01-21

[学科分类] 材料科学（综合）

[关键词] ALLOYS;CRACK PROPAGATION;DUCTILITY;EMBRITTLEMENT;FRACTURE PROPERTIES;HEAT AFFECTED ZONE;IRRADIATION;MICROSTRUCTURE;NEUTRONS;RADIATION HARDENING;RADIATIONS;REACTOR CORES;SEGREGATION;STAINLESS STEELS;STRESS CORROSION;WATER CHEMISTRY