• 한국압력기기공학회 논문집
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• 제7권1호
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• pp.35-40
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• 2011

If a plugged tube in a steam generator is broken, it may damage nearby intact tubes. To prevent this damage, it is recommended that a stabilizer is installed into the plugged tube. However, the installation cost of a stabilizer is very high. So studies are required to determine the conditions on which the installation is necessary. For this purpose static loads and dynamic loads applied on a tube should be known to estimate the residual strength and remaining fatigue and wear life of a plugged tube. Two-dimensional and three-dimensional computational fluid dynamics (CFD) analyses are performed to obtain the drag coefficient for cross flow to a tube. Using the obtained drag coefficient, the static load can be estimated and the residual strength of a plugged tube can be calculated. An inclined flow problem is also analyzed and the vertical and horizontal forces are obtained and discussed.

• 한국압력기기공학회 논문집
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• 제4권2호
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• pp.34-39
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• 2008

Steam Generator(SG) tubes are classified as one of the key components in nuclear power plants, and they should be periodically examined by the intensified management program for the assurance and diagnosis of their structural integrity. In this study, we use the optimum analysis technique to draw the detection and categorization of bowing(BOW) signals; abnormal tube-to-tube proximity in the SG upper bundle free span area. The locations in which BOW signals are detected likely have latent degradation of ODSCC(Outer Diameter Stress Corrosion Cracking). For the sake of timely and correct detection of BOW signals and diagnosis of ODSCC, we carried out the experimental demonstrations using a reduced mock-up. And we validated the MRPC(Motorized Rotating Pancake Coil) analysis technique is better than the bobbin. Hence, it comes to conclusion that the optimum analysis technique can be a good alternative for the reliable SG tube examination.

• 한국압력기기공학회 논문집
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• 제7권3호
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• pp.1-7
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• 2011

The analysis shows that the vibration is one of the main reasons of turbine failure. Especially, the problems caused by vibration occur right after retrofit of the turbine-generator and restarting the turbine. Through the case study of high vibration caused by after the turbine trip and restart, turbine vibration was identified to be influenced by startup condition. Turbine startup at high casing temperature right after unscheduled turbine trip cause radial expansion in rotor by contraction in axial direction, while casing continues to contract by steam flowing into casing. Consequently, gap between rotor and casing decrease until to metal contact to cause high vibration. Through the case study of high vibration of turbine-generator system after generator retrofit, it was identified that generator replacement could cause high vibration in turbine-generator system if the influence of generator replacement on entire system was not considered properly. To prevent startup delay caused by high vibration, it is important to keep the gaps at the design standard and start the turbine after thermal equilibrium.

• 한국압력기기공학회 논문집
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• 제7권3호
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• pp.8-14
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• 2011

The clogging of the flow area at the tube support plates(TSPs), especially at the upper TSPs results in the water level oscillation of a steam generator during normal operation. A reduction of the TSP flow area causes to increase in pressure drop within the two-phase flow zone, which destabilizes the boiling flow through the tube bundle. This phenomenon was occasionally observed at a few domestic and foreign nuclear power plants. One of the methods for defining the flow area clogging is visual inspection, which is the most effective inspection method. The results of the visual inspection for TSPs' flow area are clogging images on TSPs' quartrefoil lobes. These images are complexly distorted due to lens aberration and external factors like the distance to a subject and angle etc. In this work, we developed the analysis algorithm for clogging image of the TSP flow area of steam generators. For this purpose, we designed an image verification device applicable to the camera employed in the field for visual inspection and then, we demonstrated the validity of image analysis algorithm by using this device and commercial autoCAD program.

• 한국압력기기공학회 논문집
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• 제19권2호
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• pp.61-68
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• 2023

The phenomenon of fretting wear due to the flow-induced vibration in steam generator (SG) tube is a significant degradation mechanism in nuclear power plants. Fretting wear in SG tube is primarily attributed to the friction and impact forces between the SG tube and the tube support structures, experienced during nuclear power plants operation. While the Archard model has generally been used for the prediction of fretting wear in SG tube, it is limited by its linear nature. In this study, we introduced an "Impact Shear Work-rate" (ISW) model, which takes into account the combined effects of impact and sliding. The ISW model was evaluated using existing experimental data on fretting wear in SG tube and was compared against the Archard model. The prediction results using the ISW model were more accurate than those using the Archard model, particularly for impact forces.

• 한국압력기기공학회 논문집
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• 제19권2호
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• pp.109-116
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• 2023

This study presents a CRUD modelling to simulate the thermal resistance behavior of CRUD, deposited on the surface of a cladding tube of a fuel assembly. When heat produced from fuels transfers to a coolant through a cladding tube, the CRUD acting as an additional thermal resistance is expressed as two layers, i.e., a solid oxide layer and an imaginary fluid layer, which are added to the experimental tube's heat structure of the MARS-KS input data. The validation calculation for the experiments performed in UNIST-DISNY facility showed that the center and surface temperatures of the cladding tube increased as the porosity and the steam amount inside pores of the CRUD got higher. In addition, the temperature gradient in the imaginary fluid layer was calculated to be larger than that in the solid oxide part, indicating that the steam amount inside the layer acted more largely as thermal resistance. It was also evaluated through sensitivity calculations that the cladding tube temperature was more sensitive to the CRUD porosity and the steam amount in pores than to the inlet flow rate of the coolant.

• 한국압력기기공학회 논문집
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• 제16권1호
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• pp.42-48
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• 2020

The effect of improving the tensile and J-R fracture toughness properties of SA508 Gr.1a on the LBB margin for the main steam pipe is investigated. The material properties and microstructure images of the existing main steam piping material SA106 Gr.C used in domestic nuclear power plants and the newly selected material SA508 Gr.1a were compared. For each material, LBB margins were calculated and compared through finite element analysis and crack instability evaluation. The LBB margin of the improved SA508 Gr.1a is found to be greatly improved compared to that of the existing SA106 Gr.C and SA508 Gr.1a. This is because of the increased material's strength and J-R fracture toughness compared to the previous materials. In order to analyze the effect of physical property change on the LBB margin, the sensitivity of each LBB margin according to the variation of tensile strength and J-R fracture toughness was analyzed. The effect of the change in tensile strength was found to be greater than that of the change in fracture toughness. Therefore, an increase in strength significantly influenced the improvement of the LBB margin of the improved SA508 Gr.1a.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2014년도 추계학술대회 논문집
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• pp.394-398
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• 2014

Steam leakage is one of the major issues for the structural fracture of pipes of nuclear power plants. Therefore a method to inspect a large area of piping systems quickly and accurately is needed. Recently, the research detecting the steam leakage by using camera image is introduced. However, this method has a disadvantage. If a camera oscillates, it cannot detect a steam leakage exactly. In this paper, the technique that reduces the error due to a camera shaking is proposed. To verify the proposed method, we have performed the experiments for the oil leakage experiments. The results show that the technique can calibrate the error due to camera shaking.

• 한국소음진동공학회논문집
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• 제24권10호
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• pp.795-801
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• 2014

Steam leakage is one of the major issues for the structural fracture of pipes of nuclear power plants. Therefore a method to inspect a large area of piping systems quickly and accurately is needed. Recently, the research detecting the steam leakage by using camera image is introduced. However, this method has a disadvantage. If a camera oscillates, it cannot detect a steam leakage exactly. In this paper, the technique that reduces the error due to a camera shaking is proposed. To verify the proposed method, we have performed the experiments for the oil leakage experiments. The results show that the technique can calibrate the error due to camera shaking.

• 한국압력기기공학회 논문집
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• 제14권2호
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• pp.1-10
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• 2018

Structures, systems and components of nuclear power plants should be able to maintain safety even in the event of design-basis accidents such as high-energy line breaks. The high-pressure steam jet ejected from the broken pipe may cause damage to the adjacent structures. The ANSI/ANS 58.2 code has been adopted as a technical standard for evaluating the jet impingement load. Recently, the U.S. NRC pointed out the non-conservativeness of the ANSI/ANS 58.2, because it does not take into account the blast wave effect, dynamic behavior of the jet, and oversimplifies the shape and load characteristics of the supersonic steam jet. Therefore, it is necessary to improve the evaluation method for the high-energy line break accident. In order to evaluate the behavior of supersonic steam jet, an appropriate numerical analysis technique considering compressible flow effect is needed. In this study, numerical analysis methodology for evaluating supersonic jet impingement load was developed and verified. In addition, the conservativeness of the ANSI/ANS 58.2 model was investigated using the numerical analysis methodology. It is estimated that the ANSI jet model does not sufficiently reflect the physical behavior of under-expanded supersonic steam jet and evaluates the jet impingement load lower than CFD analysis result at certain positions.