[en] Operating experience of the various components in the nuclear power plants has shown that a variety of degradation mechanisms can occur during operation. Therefore, the accurate lifetime evaluation and systematic management are very important for the safe as well as the economical operation of the nuclear power plants. In this paper, the characteristics of a total of 17 degradation mechanisms were reviewed and the plausible degradation mechanisms such as stress corrosion cracking, fatigue, irradiation embrittlement, and so on, were identified. Also, the lifetime evaluation technologies which have been developed for the application to the domestic nuclear power plants are described. In addition, a total of 48 aging management programs which have been established for the safe operation of the various components are explained

[en] As operating time of heat exchangers progresses, fouling generated by water-borne deposits and the number of plugged tubes increase and thermal performance decreases. Both fouling and tube plugging are known to interfere with normal flow characteristics and to reduce thermal efficiencies of heat exchangers. The heat exchangers of domestic nuclear power plants have been analyzed in terms of the heat flux and heat transfer coefficient at test conditions as a means of heat exchanger management. Except for the fouling level generated in operation of heat exchangers, also, all of the tubes of heat exchangers have been replaced when the number of plugged tubes exceeds the plugging criteria based on design performance sheet. This paper describes the plugging margin evaluation method reflected the fouling of shell-and-tube heat exchangers, which can evaluate the thermal performance for heat exchangers, estimate the further fouling variations, and reflect the current fouling level. To identify the effectiveness of the developed method, the fouling and plugging margin evaluations were performed for a component cooling heat exchanger in a nuclear power plant

[en] As operating time of heat exchangers progresses, fouling generated by water-borne deposits increases and thermal performance decreases. The fouling is known to interfere with normal flow characteristics and reduce thermal efficiencies of heat exchangers. The heat exchangers of nuclear power plants have been analyzed in terms of the heat flux and heat transfer coefficient at test conditions based on the ASME OM-S/G-Part 2 as a means of heat exchanger management. It is hard to estimate the heat performance trend and to establish the future management plan. This paper describes the fouling evaluation method which can evaluate the thermal performance for heat exchangers and estimate the future fouling variations and the plugging margin evaluation method which can reflect the current fouling level developed in this study. To develop the fouling and plugging margin evaluation methods for heat exchangers, fouling factor was introduced based on the ASME O and M codes and TEMA standards. For the purpose of verifying the two evaluation methods, the fouling and plugging margin evaluations were performed for a component cooling heat exchanger in a nuclear power plant

[en] Fouling of heat exchangers is generated by water-borne deposits, commonly known as foulants including particulate matter from the air, migrated corrosion produces; silt, clays, and sand suspended in water; organic contaminants; and boron based deposits in plants. The fouling is known to interfere with normal flow characteristics and reduce thermal efficiencies of heat exchangers. This paper describes the fouling analysis technique developed in this study which can analyze the thermal performance for heat exchangers and estimate the future fouling variations. To develop the fouling analysis technique for heat exchangers, fouling factor was introduced based on the ASME O and M codes and TEMA standards. For the purpose of verifying the fouling analysis technique, the fouling analyses were performed for four heat exchangers in several nuclear power plants; two residual heat removal heat exchangers of the residual heat removal system and two component cooling water heat exchangers of the component cooling water system

[en] To perform a more realistic integrity evaluation for the dissimilar welds of NPP(Nuclear Power Plant) component, it is necessary to evaluate the microstructure and residual stress considering the actual welding process such as multi-pass welding and PWHT (Post Weld Heat Treatment). This paper proposes an integrated assessment methodology systematically developed for microstructure and residual stress on welds utilizing thermodynamics, diffusion theory, finite element method and validation experiments. And, the metallurgical microstructure and residual stress of the dissimilar weld on safety injection branch line nozzle are evaluated by application of this systematic methodology

[en] Since the suggestion of External Reactor Vessel Cooling (ERVC), the effects of malting and cooling on the response of structural integrity of the Reactor Pressure Vessel (RPV) under core melting accident conditions have been investigated. This paper describes the vessel response according to the ERVC condition and analysis method. The steady state and transient analysis for the temperature and stress field were performed using ABAQUS. Especially, transient analyses were studied for the variable boundary conditions. To obtain an analogy with real phenomena, the material properties were determined by combining and modifying the existing results considering phase transformation and temperature dependency. The results show that the vessel can be melted if there is no external cooling. Finally, the potential for vessel damage is discussed using the Larson-Miller curve and damage rule. (author)

[en] The mesh-insensitive structural stress procedure by Dong is modified to apply to the welded joints with local thickness variation and inignorable shear/normal stresses along local discontinuity surface. Validity of the modified mesh- insensitive structural stress procedure is identified comparing the structural stresses calculated for various FE models. Fatigue crack initiation cycles are determined by using the structural stresses and the various fatigue crack growth models. Fatigue test is performed to identify the validity of the fatigue analysis results. Finally, as a result of comparison between test and analysis results, it is found that the structural stress/fracture mechanics approach is valid for fatigue analysis

[en] Put cast austenitic stainless steel is used for several components, such as primary coolant piping, elbow, pump casing and valve bodies in light water reactors. These components are subject to thermal embrittlement at the reactor operating temperature. The objective of this study is to summarize the method of estimating ferrite content, Charpy impact energy and J-R curve and to evaluate the thermal embrittlement of the cast austenitic stainless steel piping used in the domestic nuclear power plants. The result of evaluation, two domestic nuclear power plants used CF-8M and CF-8A material has adequate fracture toughness after saturation

[en] During the last decade, a number of experiments and numerical analyses had been performed in conjunction with the development of simplified analytical methods to estimate the fracture behavior of cracked piping in nuclear power plant. However, the necessity of further investigation for the analytical methods was issued because of the discrepancies with the experimental data. The objective of this paper is to find out the optimum methods to evaluate the load-carrying capacities for cracked pipes. To do this, numerous analytical and finite element analyses were carried out for various pipe and crack geometries and materials. These results were synthesized for crack shapes and can be used as basic data for leak before break analyses and risk informed inspections

[en] To perform the integrity evaluation of RPV more realistically, it is necessary to evaluate the metallurgical microstructure and residual stress considering more real phenomena such as multi-pass welding process and PWHT. Accordingly, firstly, this paper proposes the integrated assessment methodology systematically developed for residual stress on weldment of RPV by using thermodynamics, diffusion theory, finite element method and validation experiment. Also, the residual stress on circumferential weldment of reactor pressure vessel is calculated considering multi-pass welding process by the commercial finite element package, ABAQUS