Abstract
IASCC has been a major concern regarding the structural and functional integrity of core internals of PWR's, especially baffle-to-former bolts. Despite numerous studies over the past few decades, additional evaluation of the parameters influencing IASCC is still needed for an accurate understanding and modeling of this phenomenon. In this study, Fe irradiation at 450 °C was used to study the cracking susceptibility of 304 L austenitic stainless steel. After 10 MeV Fe irradiation to 5 dpa, irradiation-induced damage in the microstructure was characterized and quantified along with nano-hardness measurements. After 4% plastic strain in a PWR environment, quantitative information on the degree of strain localization, as determined by slip-line spacing, was obtained using SEM. Fe-irradiated material strained to 4% in a PWR environment exhibited crack initiation sites that were similar to those that occur in neutron- and proton-irradiated materials, which suggests that Fe irradiation may be a representative means for studying IASCC susceptibility. Fe-irradiated material subjected to 4% plastic strain in an inert argon environment did not exhibit any cracking, which suggests that localized deformation is not in itself sufficient for initiating cracking for the irradiation conditions used in this study.
Highlights
Austenitic stainless steel, owing to its high strength, ductility, and fracture toughness, has been selected as the structure alloy for the majority of the core internals of the Pressurized Water Reactor (PWR)
The core internal components made of stainless steels have been shown to experience Irradiation Assisted Stress Corrosion Cracking (IASCC), which may affect the integrity of the PWRs
The mean size of Frank loops obtained was similar to the value (12.5 nm) reported in Ref. [34] on 304 grade stainless steel irradiated with 2.8 MeV Fe at 300 C at 10 dpa, and the number density was about a factor 10 lower in this study which may be due to the higher irradiation temperature, which tends to decrease the defect density as reported in Ref. [35]
Summary
Austenitic stainless steel, owing to its high strength, ductility, and fracture toughness, has been selected as the structure alloy for the majority of the core internals of the Pressurized Water Reactor (PWR). The core internal components made of stainless steels have been shown to experience Irradiation Assisted Stress Corrosion Cracking (IASCC), which may affect the integrity of the PWRs. Core internals are in close proximity to the core and face significant neutron irradiation of up to 80 dpa during the designed lifetime of a PWR. Core internals are in close proximity to the core and face significant neutron irradiation of up to 80 dpa during the designed lifetime of a PWR This irradiation damage has a devastating influence on the material as it modifies the microstructure (by inducing defects such as dislocation loops, precipitates, cavities, etc.) and the mechanical properties (such as tensile properties, ductility, and fracture toughness) [1e4].
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