The original fracture criteria developed by Maxey/Kiefner for axial through-wall and surface-cracked pipes have worked well for many industries for a large variety of relatively low strength and toughness materials. However, newer line pipe steels have some unusual characteristics that differ from these older materials. One example is a test data that has demonstrated that X80 line-pipe with an axial through-wall-crack can fail at pressures about 30 percent lower than predicted with commonly used analysis methods for older steels. Thus, it is essential to review the currently available models and investigate the applicability of these models to newer high-strength line pipe materials. In this paper, the available models for predicting the failure behavior of axial-cracked pipes (through-wall-cracked and external surface-cracked pipes) were reviewed. Furthermore, the applicability of these models to high-strength steel pipes was investigated by analyzing limited full-scale pipe fracture initiation test results. Based on the analyzed results, the shortcomings of the available models were identified. For both through-wall and surface cracks, the major shortcomings were related to the characterization of the material toughness, which generally leads to non-conservative predictions in the J-T analyses. The findings in this paper may be limited to the test data that were consider for this study. The requisite characteristics of a potential model were also identified in the present paper.

The thermal stratification phenomenon in the nuclear power plant can cause abnormal deformation of the piping, contact with the support, damage to the support system. Repetition of the thermal stratification phenomenon or variation of the thermal boundary layer can cause thermal fatigue. Thermal stratification phenomenon in nuclear power plants is still an ongoing issue and active research has been carried out. In this paper, the current situation in Korean nuclear power plants is described, followed by the status of research and the future problems on the thermal stratification phenomenon in Korea.

This paper presents a method to quantify R-ratio effect on J-integral under large-amplitude cyclic loading condition. Generally, monotonic tearing resistance curves are used to assess stability of cracked nuclear piping under seismic loading. However, it is well known that fracture toughness decrease at cyclic loading condition, according to R-ratio. For this reason, it is important to quantify the J-R curves under cyclic loading condition. To quantify the R-ratio effect, correction method which was proposed by Tranchand is considered. This method considers crack opening area in order to calculate modified J-integral. This method leads to an increase of fracture toughness. At R=-0.5 case, this method is good agreement with monotonic J-R curves. However, results show that this method has a limit to apply a large R-ratio case.

Flow accelerated corrosion (FAC) of the carbon steel piping has been a significant problem in nuclear power plants. FAC occurs under certain hydrodynamic, environmental, and material conditions, and extensive research into the factors of FAC has been conducted. The basic process of FAC is now relatively well understood; however, a full mechanistic model has not yet been established. Recently, the Korea Atomic Energy Research Institute (KAERI) has built a large experiment loop system for FAC. To produce significant experimental results using this system, the factors affecting on FAC should be analyzed quantitatively, and a model needs to be developed. In this work, a statistical modeling methodology to develop an empirical model is described in detail, and a preliminary model is suggested. Firstly, FAC data were collected from the research literature in Japan and the results of domestic experiments. The flow rate, water temperature, pH at room temperature, and the Cr content are selected as major factors, and nonlinear regression is used to find the best fit of the available data. An iterative procedure between suggesting and evaluating a model is used until an optimum model is obtained. The developed model gives the FAC rate comparable to the measured FAC rate. The developed model is going to be refined using additional laboratory data in the future.

The study presents structural integrity assessment of ductility degradation of a baffle former assembly by performing finite element analysis considering real loading conditions and stress triaxiality. Variations of fracture strain curves of type 304 austenitic stainless steel with stress triaxiality are derived based on the previous study results. Temperature distributions during normal operation such as heat-up, steady state, and cool-down are calculated via finite element temperature analysis considering gamma heating and heat convection with reactor coolant. Variations of stress and strain state during long operation period are also calculated by performing sequentially coupled temperature-stress analysis. Fracture strain is derived by using the fracture curve and the stress triaxility. Finally, variations of ductility degradation damage indicator with the fracture strain and the equivalent inelastic strain are investigated. It is found that maximum value of the ductility degradation damage index continuously increases and becomes 0.4877 at 40 EFPYs. Also, the maximum value occurs at top and middle inner parts of the baffle former assembly before and after 20 EFPYs, respectively.

Since, in case of high energy piping, steam jets ejected from the rupture zone may cause damage to nearby structure, it is necessary to design it into consideration of nuclear power plant design. For the existing nuclear power plants, the ANSI / ANS 58.2 technical standard for high-energy pipe rupture was used. However, the US Nuclear Regulatory Commission (USNRC) and academia recently have pointed out the non-conservativeness of existing high energy pipe fracture evaluation methods. Therefore, it is necessary to develop a highly reliable evaluation methodology to evaluate the behavior of steam jet ejected during high energy pipe rupture and the effect of steam jet on peripheral devices and structures. In this study, we develop a method for analyzing the impact load of a jet by high energy pipe rupture, and plan to carry out an experiment to verify the evaluation methodology. In this paper, the basic data required for the design of the jet impact load experiment equipment under construction, 1) the load change according to the jet distance, 2) the load change according to the jet collision angle, 3) the load variation according to structure diameter, and 4) the load variation depending on the jet impact position, are numerically obtained using the developed steam jet analysis technique.

Following the 2011 Fukushima Nuclear Power Plant accident, the IAEA has issued a revised version of the Nuclear Safety Standard for beyond design basis earthquake to consider the core meltdown accident. In Korea, relevant laws and regulations were also revised to consider beyond design basis earthquake to nuclear components. In this paper, CAV, an seismic damage factor that determines the restart of nuclear power plant after operating breakdown earthquake, is proposed for extension to the beyond design basis earthquake. For pipings not satisfying the beyond design basis earthquake condition, several evaluation methods are suggested, such as strain-based evaluation methods, simple nonlinear analysis method and cumulative damage evaluation method.