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Few studies on social welfare in South Korea have examined the effects of climate change, especially heat waves on vulnerable populations. The present study aims to investigate how heat waves affect vulnerable populations. This study utilized a cross-sectional study design, using the daily heat index and heat-related mortality data for Seoul, South Korea, in summer 2018. The research used micro-raw data of deaths caused by hypertensive, ischemic heart, and cerebrovascular diseases, as well as heat index data. An effect was observed for the heat index on mortality for individuals over 65 years of age, men, people with spouses, and those ages 75–79 years. [ABSTRACT FROM AUTHOR]

This study proposes a methodology to estimate time-varying in situ wear coefficient between steam generator tubes in nuclear power plants and their supporting structures. Actual wear depth measurement data of steam generator tubes of OPR1000 (Optimized Power Reactor 1000 MW) plants in Korea were collected and analyzed to investigate the behavior of fretting wear. To determine the in situ wear coefficient, a mathematical expression was developed as a function of various parameters such as measured wear depth time history, work rate, contact geometry of the tube, and its support. These calculated in situ wear coefficients were then used to obtain wear depth history curves. Results obtained were then compared with actual field measurement data to show the validity of the proposed method. Many researchers have obtained wear coefficients under laboratory conditions. However, those coefficients cannot be considered as realistic factors for operating steam generators. The in situ wear coefficient proposed in this study is based on wear measurement data obtained from real operating steam generator tubes. Therefore, they can be used to precisely predict the wear depth of steam generator tubes, thus allowing safe and economical management of steam generators. [ABSTRACT FROM PUBLISHER]

Abstract: Flow induced vibration of steam generator tubes potentially causes excessive fretting wear at the supports such as anti-vibration bars and tube support plates. For a reliable design of tubes against the flow-induced vibration related failure, the prediction of vibration and wear of tubes should be performed through complicated steps including the thermal-hydraulic analysis, dynamic modal analysis, evaluation of fluid-elastic instability, prediction of turbulence-induced vibration and wear depth for thousands of tubes. However, entire tubes cannot be evaluated within a limited time of design engineering by the conventional analysis methodology. In this paper, we describe an efficient computer program to assess the structural integrity of steam generator tubes against the flow-induced vibration related failure in a very systematic way. The program contains all the necessary thermal-hydraulic database of typical steam generators. It has a very special function to perform modal analysis for all thousands of tubes of a steam generator much faster than the conventional method. The program also performs fluid-elastic instability analysis and calculates the vibrational response to the turbulent flow excitation, and then can predict the wear depth for all tubes of a steam generator. Finally, we can generate the wear prediction map for whole tubes so that an efficient and practical steam generator maintenance management program is feasible. The utilization of the developed computer program for the design and maintenance of steam generators can significantly increase the structural integrity of steam generator tubes of a nuclear power plant. [Copyright &y& Elsevier]

The present paper addresses the mechanical characteristics of the stainless steel narrow gap weld (NGW) for the leak-before-break (LBB) application to the main loop piping of a nuclear power plant. Recently, in Korea, the connection with the reactor coolant main loop piping and the steam generator has been welded with ER308L NGW after the replacement of a steam generator of a Korean nuclear power plant. The NGW technique has many merits, for instance, the reduction of construction time and the reduction of shrinkage and deformation after welding due to its small groove angle and welding bead width compared with the conventional welds. In this paper, the tensile and fracture toughness test results of the three ER308L test coupons from NGW were presented and compared with those from conventional welds at the operating condition of the nuclear power plant. In addition, the distribution of the welding residual stress as well as the deformation behavior of the ER308L weld due to NGW was predicted through the nonlinear two-dimensional finite element analysis in which the detailed actual welding process of NGW was simulated. The results presented in this paper can be used to evaluate LBB application to nuclear piping with NGW and to provide the important information to perform the flaw evaluation as well as improve the weld procedure of NGW. [ABSTRACT FROM AUTHOR]

Abstract: A typical pressurized water reactor (PWR) steam generator has approximately 10,000 tubes. These tubes have different geometries, supporting conditions, and different material properties due to the non-uniform temperature distribution throughout the steam generator. Even though some tubes may have the same geometry and boundary conditions, the non-uniform distribution of coolant densities adjacent to the tubes causes them to have different added mass effects and dynamic characteristics. Therefore, for a reliable design of the steam generator, a separate modal analysis for each tube is necessary to perform the FIV (flow-induced vibration) analysis. However, the modal analysis of a tube including the finite element modeling is cumbersome and takes lots of time. And when a commercial finite element code is used, interfacing the modal analysis result, such as natural frequencies and mode shapes, with the FIV analysis procedure requires an additional significant amount of time and can possibly incur inadvertent error due to the complexity of data processing. It is therefore impossible to perform the complete FIV analysis for ten thousands of tubes when designing or maintaining a steam generator although it is necessary. Rather, to verify the safe design against the FIV, only a couple of tubes are chosen based on engineering judgment or past experience. In this paper, a computer program, PIAT-MODE, was developed which is able to perform modal analysis of all tubes of a PWR steam generator in a very efficient way. The geometries and boundary conditions of every tube were incorporated into PIAT-MODE using appropriate mathematical formulae. Material property data including the added mass effect was also included in the program. Once a specific tube is selected, the program automatically constructs the finite element model and generates the modal data very quickly. Therefore, modal analysis can be performed for every single tube in a straight way. When PIAT-MODE is coupled with the FIV analysis program, it gives an amazing benefit, which makes the FIV analysis of all tubes possible within a limited time during the design or maintenance period. Using the developed program, the stability ratio regarding the fluid-elastic instability and the amplitude of vibration resulting from the turbulence flow excitation can be calculated for all tubes according the standard ASME Code, therefore, much more reliable design of the steam generator against the FIV related failures can be achieved. For an operating plant, there is a requirement that every single tube must be quantitatively checked whether wear would be more than 40% in thickness during the next operation period. If yes, the tube must be plugged to prevent severe failure including the tube wall penetration. In order to decide the tube plugging, which leads the plant performance degradation, the wear prediction is required. The wear prediction analysis needs wear data by ultrasonic testing and modal property data of tubes. PIAT-MODE will be a power tool enabling the wear prediction in a limited maintenance period by supplying modal data for all tubes. [Copyright &y& Elsevier]