• 대한기계학회논문집A
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• 제27권5호
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• pp.724-733
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• 2003

Fatigue design rules for welds in the ASME Boiler and Pressure Vessels Code are based on the use of Fatigue Strength Reduction Factors(FSRF) against a code specified fatigue design curve generated from smooth base metal specimens without the presence of welds. Similarly, stress intensification factors that are used in the ASME B3l.1 Piping Code are based on component S-N curves with a reference fatigue strength based on straight pipe girth welds. But the determination of either the FSRF or stress intensification factor requires extensive fatigue testing to take into account the stress concentration effects associated with various types of component geometry, weld configuration and loading conditions. As the fatigue behavior of welded joints is being better understood, it has been generally accepted that the difference in fatigue lives from one type of weld to another is dominated by the difference in stress concentration. However, general finite element procedures are currently not available for effective determination of such stress concentration effects. In this paper, a mesh-insensitive structural stress method is used to re-evaluate the S-N test data, and then more effective method is proposed for pressure vessel and piping fatigue design.

• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.56-61
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• 2013

This paper presents the results of integrity assessment for the cracks happened in reactor vessel upper head penetration nozzles. The crack morphology for a boat sample from crack area was analyzed through microscope. The stress condition including weld residual stress around crack was analyzed using finite element analysis. From the results of crack morphology and stress condition, the crack was concluded as primary water stress corrosion cracking. The integrity of the cracked nozzle was assessed by the methodology provided in ASME Section XI. According to the assessment results, the remaining life of the cracked nozzle was 1.43 yrs. and the plant decided to repair it.

• 한국압력기기공학회 논문집
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• 제19권2호
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• pp.140-145
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• 2023

When performing stress analysis for a nozzle in nuclear power plants, the nozzle loads should be determined conservatively. Existing stress analysis report of 3-D nozzle loads in nuclear power plants often provide only load magnitude not the sign (direction). Since calculated stress distribution depends on load direction, determining critical load directions for conservative stress analysis is crucial. In this study, an efficient method for determining critical load directions in nozzle loads is proposed. In the proposed method, stresses are firstly calculated using elastic finite element (FE) analysis for the uni-axial load in each direction. Then stress distributions for the multi-axial loads are analytically calculated using the principle of superposition. The calculated stress values are verified by comparing with FE analysis results under multi-axial loading. By using this method, the complex task of determining conservative load directions can be simplified.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3361-3379
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• 2022

A nuclear power plant (NPP) piping is designed against low-frequency earthquakes. However, earthquakes that can occur at NPP sites in the eastern part of the United States, northern Europe, and Korea are high-frequency earthquakes. Therefore, this study conducts bi-directional shaking table tests on actual-scale NPP piping and studies the response characteristics of low- and high-frequency earthquake motions. Such response characteristics are analyzed by comparing several responses that occur in the piping. Also, based on the test results, a piping numerical analysis model is developed and validated. The piping seismic performance under high-frequency earthquakes is derived. Consequently, the high-frequency excitation caused a large amplification in the measured peak acceleration responses compared to the low-frequency excitation. Conversely, concerning relative displacements, strains, and normal stresses, low-frequency excitation responses were larger than high-frequency excitation responses. Main peak relative displacements and peak normal stresses were 60%-69% and 24%-49% smaller in the high-frequency earthquake response than the low-frequency earthquake response. This phenomenon was noticeable when the earthquake motion intensity was large. The piping numerical model simulated the main natural frequencies and relative displacement responses well. Finally, for the stress limit state, the seismic performance for high-frequency earthquakes was about 2.7 times greater than for low-frequency earthquakes.

• 한국정밀공학회지
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• 제19권6호
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• pp.192-199
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• 2002

Largely, there are three kinds of the design criteria of piping system in petrochemical plant. The first is on the pipe thickness in accordance with the design pressure of piping system. The second is on the static state evaluation by thermal growth and the other is on the dynamic evaluation by piping vibration. According to the ASME B31.3 code, the internal pressure design thickness fur straight pipe shall be calculated as a code formula. And the static design by thermal displacement is defined 7000 cycles of fatigue life in operating the piping system with a design condition. However, the dynamic design evaluation in comparative with small displacements of high frequencies to the static condition has not established clearly the method, yet. So, this study purposes to present the trial of a proposal of dynamic design criterion on the basis of static design method.

• 한국압력기기공학회 논문집
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• 제14권1호
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• pp.58-65
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• 2018

The purpose of this study is to develop a Half Nozzle Repair(HNR) process to prevent the leakage from welds on small bore piping in Reactor Coolant System. The Codes & Standards of tempered bead and design requirements of J-Groove welds are reviewed. Automatic machine GTAW welding and machining equipments are developed to perform HNR process. Single pass welding and overlay welding equipments are conducted in order to obtain the optimal temper bead welding process parameters with Alloy 52M filler wire. Coarse grain heat affected zone(CGHAZ) is formed by rapid cooling rate in heat affected zone after welding. Accordingly, a proper temper bead technique is required to reduce CGHAZ in 1-Layer of welds by 2- and 3-Layers. Mock-up tests show that the developed HNR process is possible to meet ASME Code & Standard requirements without any defect.

• 대한기계학회논문집B
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• 제33권5호
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• pp.381-388
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• 2009

In some portions of nuclear piping systems, stratification phenomena may occur due to the density difference between hot and cold stream. When the temperature difference is large, the stratified flow under diverse operating conditions can produce high thermal stress, which leads to unanticipated piping integrity issues. The objectives of this research are to examine controvertible numerical factors such as model size, grid resolution, turbulent parameters, governing equation, inflow direction and pipe wall. Parametric three-dimensional computational fluid dynamics analyses were carried out to quantify effects of these parameters on the accuracy of temperature profiles in a typical nuclear piping with complex geometries. Then, as a key finding, it was recommended to use optimized mesh of real piping with the conjugated heat transfer condition for accurate thermal stratification analyses.

• Nuclear Engineering and Technology
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• 제45권2호
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• pp.237-248
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• 2013

Thermal stratification has continuously caused several piping failures in nuclear power plants since the early 1980s. However, this critical thermal effect was not considered when the old nuclear power plants were designed. Therefore, it is urgent to evaluate this unexpected thermal effect on the structural integrity of piping systems. In this paper, the thermal effects of stratified flow in two different safety injection piping systems were investigated by using a coupled CFD-FE method. Since stratified flow is generally generated by turbulent penetration and/or valve leakage, thermal stress analyses as well as CFD analyses were carried out considering these two primary causes. Numerical results show that the most critical factor governing thermal stratification is valve leakage and that temperature distribution significantly changes according to the leakage path. In particular, in-leakage has a high possibility of causing considerable structural problems in RCS piping.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2005년도 춘계학술대회 논문집
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• pp.1303-1306
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• 2005

The mechanical properties of bellows, such as the extensibility and the strength can be changed depending on the shape. For the shipbuilding material, it is favorable that the fatigue life is long due to the elastic property and the reduction of thermal stress in piping system. Nowadays, the domestic production and design of bellows are based on the E.J.M.A Code. Therefore, the design standard is in need because of much errors and lack of detailed analysis. In this study, it is attempted to find out the optimal shape of U-type bellows using the finite element method. The effective factors, mountain height, length, thickness, and number of mountains and the length of joint are considered and the proper values are chosen for the simulation. The results shows that if the number of mountains are reduced, the volume decreases while the stress increases. However, the number of mountains are increased, the volume increases above the standard volume and the stress obviously increases. In addition, the effect of the thickness of bellows on the stress is very large. Both of the volume and stress are decreasing at a certain lower value region.

• 한국압력기기공학회 논문집
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• 제8권3호
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• pp.7-12
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• 2012

Steam generator tubes are critical boundary of the primary and secondary side in nuclear power plants. Eddy current testing is commonly used as the method of non-destructive testing for the safety and integrity of steam generator tubes in the nuclear power plants. Changes in the geometric shape act as a stress concentration factor likely to cause a defect during the steam generator operation. The mixed-signals with the geometric shape are distorted and attributes that are difficult to detect signals. An example is bending stress due to compression process at a U-bend occurring in the intrados region which has a small radius of curvature. The resulting change in the geometric shape may lead to a dent like occurrences. The dent can cause stress concentration and generates stress corrosion cracks. In this study, the steam generator tubes of nuclear power plant were selected to study for analysis of mixed-signal containing dent and stress corrosion cracks.