Abstract
The performance of structural materials is critical for the safe and economic operation of light water reactors. During power operation, reactor core internal materials are exposed to aggressive corrosive coolant environment, vigorous thermal/mechanical loading, and intensive neutron irradiation. Such severe service conditions can activate and enhance a wide range of degradation processes, leading to deteriorated material properties and service performance. To ensure the structural integrity and functionality of nuclear reactor components, material degradation and damage mechanisms must be understood and managed adequately. It has been recognized that there are knowledge and data gaps in the existing information and technical bases for long-term operation and aging management. In particular, post-irradiation data on fracture toughness and crack growth rate are lacking. In this work, irradiated materials harvested from a decommissioned reactor are studied for their cracking susceptibility and fracture resistance as a function of irradiation dose. The materials are a Type 304 stainless steel sectioned from the baffle plates of a pressurized water reactor after 38 years of service. Miniature compact-tension specimens about 6.5-mm thick are machined from these materials with different levels of irradiation damage, ranging from < 1 to ∼50 dpa depending on the original locations with respect to the reactor core. Crack growth rate and fracture toughness J-R curve tests are performed in a low-corrosion-potential environment at ∼315°C. All samples behave similarly under cyclic loading, and no deteriorated corrosion fatigue behavior can been seen in the test environment. Under constant loads, most of samples show no elevated crack growth rates, suggesting an adequate stress corrosion cracking resistance for these irradiated samples in the test environment. An unstable cracking behavior was observed occasionally where step-wise crack advances upon load increases can be seen. The effect of neutron irradiation is evident on fracture toughness. With the increasing dose, the J-R curve declined constantly, and became very shallow at high doses. It is evident that this baffle plate material has been severely embrittled by neutron irradiation. In addition, an unexpected fully IG morphology has been observed for the all high-dose samples fractured at room temperature in air atmosphere. The occurrence of this brittle fracture in the absence of aggressive environment confirmed a high degree of embrittlement of this material resulting from its service exposure to neutron irradiation.
Electrochemical Corrosion Potential Modeling in the Primary Heat Transport Circuit of the Chinshan Boiling Water Reactor under the Condition of Hydrogen Water Chemistry with Noble Metal Coating