Effects of Thermal Aging and Neutron Irradiation on Crack Growth Rate and Fracture Toughness of Cast Stainless Steels and Austenitic Stainless Steel Welds

2014 ◽  
Author(s):  
Omesh K. Chopra
Keyword(s):  
Fracture Toughness ◽  
Growth Rate ◽  
Stainless Steel ◽  
Crack Growth Rate ◽  
Austenitic Stainless Steel ◽  
Crack Growth ◽  
Neutron Irradiation ◽  
Stainless Steels ◽  
Thermal Aging ◽  
Steel Welds

scholarly journals Fracture Resistance of Cast Austenitic Stainless Steels

2016 ◽  
Author(s):  
Y. Chen ◽  
W-Y. Chen ◽  
A. S. Rao ◽  
Z. Li ◽  
Y. Yang ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Growth Rate ◽  
Corrosion Resistance ◽  
Neutron Irradiation ◽  
Stainless Steels ◽  
Thermal Aging ◽  
Austenitic Stainless Steels ◽  
Light Water Reactor ◽  
Delta Ferrite ◽  
Light Water

Cast austenitic stainless steels (CASS) possess excellent corrosion resistance and mechanical properties and are used alongside with wrought stainless steels (SS) in light water reactors for primary pressure boundaries and reactor core internal components. In contrast to the fully austenitic microstructure of wrought SS, CASS alloys consist of a dual-phase microstructure of delta ferrite and austenite. The delta ferrite is critical for the service performance since it improves the strength, weldability, corrosion resistance, and soundness of CASS alloys. On the other hand, the delta ferrite is also vulnerable to embrittlement when exposed to reactor service temperatures and fast neutron irradiations. In this study, the combined effect of thermal aging and neutron irradiation on the degradation of CASS alloys was investigated. Neutron-irradiated CASS specimens with and without prior thermal aging were tested in simulated light water reactor environments for crack growth rate and fracture toughness. Miniature compact-tension specimens of CF-3 and CF-8 alloys were tested to evaluate the extent of embrittlement resulting from thermal aging and neutron irradiation. The materials used are static casts containing more than 23% delta ferrite. Some specimens were thermally aged at 400 °C for 10,000 hours prior to the neutron irradiation to simulate thermal aging embrittlement. Both the unaged and aged specimens were irradiated at ∼320°C to a low displacement damage dose of 0.08 dpa. Crack growth rate and fracture toughness J-integral resistance curve tests were carried out on the irradiated and unirradiated control samples in simulated light water reactor environments with low corrosion potentials. While no elevated crack propagation rates were detected in the test environments, significant reductions in fracture toughness were observed after either thermal aging or neutron irradiation. The loss of fracture toughness due to neutron irradiation seemed more evident in the samples without prior thermal aging. Transmission electron microscope (TEM) examination was carried out on the thermally aged and neutron irradiated specimens. The result showed that both neutron irradiation and thermal aging can induce significant changes in the delta ferrite. A high density of G-phase precipitates was observed with TEM in the thermally aged specimens, consistent with previous results. Similar precipitate microstructures were also observed in the neutron-irradiated specimens with or without prior thermal aging. A more extensive precipitate microstructure can be seen in the samples subjected to both thermal aging and neutron irradiation. The similar precipitate microstructures resulting from thermal aging and neutron irradiation are consistent with the fracture toughness results, suggesting a common microstructural origin of the observed embrittlement after thermal aging and neutron irradiation.

Cyclic Crack Growth Rate of Irradiated Austenitic Stainless Steel Welds in Simulated BWR Environment

2011 ◽  
Author(s):  
Y. Chen ◽  
B. Alexandreanu ◽  
W. J. Shack ◽  
K. Natesan ◽  
A. S. Rao
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Crack Growth ◽  
Neutron Irradiation ◽  
Growth Rates ◽  
Reactor Core ◽  
Cracking Behavior ◽  
Steel Welds ◽  
The Impact ◽  
Crack Growth Rates

Reactor core internal components in light water reactors are subjected to neutron irradiation. It has been shown that the austenitic stainless steels used in reactor core internals are susceptible to stress corrosion cracking after extended neutron exposure. This form of material degradation is a complex phenomenon that involves concomitant conditions of irradiation, stress, and corrosion. Interacting with fatigue damage, irradiation-enhanced environmental effects could also contribute to cyclic crack growth. In this paper, the effects of neutron irradiation on cyclic cracking behavior were investigated for austenitic stainless steel welds. Post-irradiation cracking growth tests were performed on weld heat-affected zone specimens in a simulated boiling water reactor environment, and cyclic crack growth rates were obtained at two doses. Environmentally enhanced cracking was readily established in irradiated specimens. Crack growth rates of irradiated specimens were significantly higher than those of nonirradiated specimens. The impact of neutron irradiation on environmentally enhanced cyclic cracking behavior is discussed for different load ratios.

Cracking Behavior of a Decommissioned Material in Light Water Reactor Environment

2020 ◽  
Author(s):  
Y. Chen ◽  
B. Alexandreanu ◽  
A. S. Rao
Keyword(s):  
Fracture Toughness ◽  
Growth Rate ◽  
Crack Growth Rate ◽  
Crack Growth ◽  
Neutron Irradiation ◽  
Reactor Core ◽  
Test Environment ◽  
Cracking Behavior ◽  
Light Water ◽  
Water Reactor

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.

SCC Growth Rates and Reference Curves for Low Carbon Stainless Steels in BWR Environment

2004 ◽  
Author(s):  
Mikiro Itow ◽  
Masaaki Kikuchi ◽  
Norihiko Tanaka ◽  
Jiro Kuniya ◽  
Michiyoshi Yamamoto ◽  
...  
Keyword(s):  
Growth Rate ◽  
Stainless Steel ◽  
Crack Growth Rate ◽  
Crack Growth ◽  
Stainless Steels ◽  
Nuclear Power ◽  
Low Carbon ◽  
Operation And Maintenance ◽  
System Components ◽  
Reference Curves

Nuclear Plant Operation and Maintenance Code has been developed and is going to be applied for nuclear power system components in Japan. If a crack is detected in a component, the evaluation of crack growth due to stress corrosion cracking (SCC) is required. In recent years, the components in BWR primary systems made of low carbon stainless steel, such as core shroud and PLR piping, have suffered from SCC and it is necessary to prepare the crack growth rate reference curves for the materials. In this paper, the development of the SCC growth rate database for low carbon stainless steel in BWR water and the proposed reference curves in Japan are described.

scholarly journals Crack Growth Rate and Fracture Toughness Tests on Irradiated Cast Stainless Steels

2013 ◽  
Author(s):  
Y. Chen ◽  
B. Alexandreanu ◽  
K. Natesan
Keyword(s):  
Fracture Toughness ◽  
Growth Rate ◽  
Crack Growth Rate ◽  
Crack Growth ◽  
Stainless Steels
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