Retardation Effect of Cyclic Overload on Stress Corrosion Crack Growth in Stainless Steel

2017 ◽  
Author(s):  
Toshiyuki Saito ◽  
Takahiro Hayashi ◽  
Chihiro Narazaki ◽  
Mikiro Itow
Keyword(s):  
Stress Corrosion ◽  
Stainless Steels ◽  
Power Plants ◽  
Structural Integrity ◽  
Plastic Region ◽  
Austenitic Stainless Steels ◽  
Growth Behavior ◽  
Welding Residual Stress ◽  
Earthquake Force ◽  
Cyclic Overload

Stress Corrosion Cracking (SCC) has been observed in some components of austenitic stainless steels in the Boiling Water Reactors (BWRs). The structural integrity evaluation for flawed component is performed for continued service for a specified time period based on the Rules on Fitness-for-Service (FFS) for Nuclear Power Plants, such as JSME FFS Code or ASME Section XI. SCC growth evaluation is generally performed only by taking into account steady loads, such as welding residual stress. It is important to examine various factors affecting SCC growth behavior for further understanding and improvement in predicting growth behavior in the BWR environment. Cyclic overloading due to such as earthquake force is one of the important factors to be evaluated. In this study, the effect of cyclic overload on SCC growth in simulated BWR environment has been examined by using CT specimens of cold-rolled stainless steels (Type 316L). The retardation phenomenon was observed in SCC growth behavior immediately after the cyclic overloading was applied. It was considered that SCC propagation was retarded due to the compressive plastic region at the crack tip, introduced by overloads. The method of predicting the SCC growth behavior after cyclic overloading was also discussed.

Fracture Toughness Criteria of Irradiated Austenitic Stainless Steels for Structural Integrity Evaluation of BWR Internal Components

2019 ◽  
Author(s):  
Takahiro Hayashi ◽  
Shigeaki Tanaka ◽  
Tomonori Abe ◽  
Seiji Sakuraya ◽  
Suguru Ooki ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Stainless Steels ◽  
Power Plants ◽  
Structural Integrity ◽  
Neutron Fluence ◽  
Austenitic Stainless Steels ◽  
Evaluation Methodology ◽  
Irradiation Damage ◽  
Materials Testing ◽  
Testing Data

Abstract Continuous improvement of the structural integrity evaluation methodology in the plant life management (PLM) evaluations is of increasing importance for aged light water reactors. In PLM evaluations, structural integrity evaluations are required for degradation mechanisms considered in the subject equipment and components. Austenitic stainless steels used in reactor internal components are known to show decreases in ductility and fracture toughness due to accumulated neutron irradiation damage. In Japan, “Rules on Fitness-for-Service for Nuclear Power Plants of the Japan Society of Mechanical Engineers Code (JSME FFS Code)” provides fracture evaluation method and criterion, based on the linear elastic fracture mechanics, for irradiated stainless steels of boiling water reactor (BWR) internal components. The fracture toughness criterion, however, was developed with limited materials testing data and knowledge available at that time and it has not been revised since the code originally established. In this study, fracture toughness criteria for structural integrity evaluation were discussed and developed with the latest database on fracture toughness of irradiated austenitic stainless steels, including additional material testing data obtained in this study for the neutron fluence range of interest from 1 to 3 dpa. First, the fracture toughness data of austenitic stainless steels irradiated in BWR conditions were compiled to evaluate the correlation between fracture toughness and neutron fluence. Material characteristics potentially affecting fracture toughness, such as chemical composition and specimen orientation, were also considered and discussed in the development of the fracture toughness criteria. Based on the results, the fracture toughness criteria for irradiated austenitic stainless steels were proposed for fracture evaluation of the BWR internal components.

Structural Integrity Research for Reactor Pressure Vessel Under In-Vessel Retention of a Core Melt

2016 ◽  
Author(s):  
Yongjian Gao ◽  
Yinbiao He ◽  
Ming Cao ◽  
Yuebing Li ◽  
Shiyi Bao ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Material Properties ◽  
Power Plants ◽  
Structural Integrity ◽  
Plastic Material ◽  
Plastic Region ◽  
Reactor Pressure Vessel ◽  
Elastic Plastic ◽  
Elastic Plastic Material ◽  
Reactor Pressure

In-Vessel Retention (IVR) is one of the most important severe accident mitigation strategies of the third generation passive Nuclear Power Plants (NPP). It is intended to demonstrate that in the case of a core melt, the structural integrity of the Reactor Pressure Vessel (RPV) is assured such that there is no leakage of radioactive debris from the RPV. This paper studied the IVR issue using Finite Element Analyses (FEA). Firstly, the tension and creep testing for the SA-508 Gr.3 Cl.1 material in the temperature range of 25°C to 1000°C were performed. Secondly, a FEA model of the RPV lower head was built. Based on the assumption of ideally elastic-plastic material properties derived from the tension testing data, limit analyses were performed under both the thermal and the thermal plus pressure loading conditions where the load bearing capacity was investigated by tracking the propagation of plastic region as a function of pressure increment. Finally, the ideal elastic-plastic material properties incorporating the creep effect are developed from the 100hr isochronous stress-strain curves, limit analyses are carried out as the second step above. The allowable pressures at 0 hr and 100 hr are obtained. This research provides an alternative approach for the structural integrity evaluation for RPV under IVR condition.

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