• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2002년도 추계학술대회논문집
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• pp.921-926
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• 2002

In this study, we investigate the plugging effect on the CE type steam generator tube. The natural frequency and mode shape will be changed due to decrease of the effective mass distribution along the tube. We compared the variation of stability ratio for plugged tube with that for unplugged one. The natural frequency increased because of removing the cooling water inside the steam generator tube, but the stability ratio decreased inversely because of changing the vibrational model shape. We also investigated the turbulent excitation effect.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2002년도 추계학술대회논문초록집
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• pp.390.1-390
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• 2002

In this study, we investigate the plugging effect on the CE type steam generator tube. The natural frequency and mode shape will be changed due to decrease of the effective mass distribution along the tube. We compared the variation of stability ratio for plugged tube with that fur unplugged one. The natural frequency increased because of removing the cooling water inside the steam generator tube, but the stability ratio decreased inversely because of changing the vibrational mode shape. We also investigated the turbulent excitation effect.

• 한국소음진동공학회논문집
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• 제12권4호
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• pp.261-271
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• 2002

The fluid-elastic instability analysis of the U-tube bundle inside the steam generator is very important not only for detailed design stage of the SG but also for the change of operating condition of the nuclear powerplant. However the calculation procedure for the fluid-elastic instability was so complicated that the consolidated computer program has not been developed until now. In this study, the numerical calculation procedure and the computer program to obtain the stability ratio were developed. The thermal-hydraulic data in the region of secondary side of steam generator was obtained from executing the ATHOS3 code. The distribution of the fluid density can be calculated by using the void fraction, enthalpy, and operating pressure. The effective mass distribution along the U-tube was required to calculate natural frequency and dynamic mode shape using the ANSYS ver. 5.6 code. Finally, stability ratios for selected tubes of the CE type steam generator were computed. We considered the YGN 3.4 nuclear powerplant as the model plant, and stability ratios were investigated at the flow exit region of the U-tube. From our results, stability ratios at the central and the outside region of the tube bundle are much higher than those of other region.

• 한국음향학회지
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• 제43권1호
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• pp.122-130
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• 2024

에어컨 실외기 내부의 압축기 진동 소음은 실외기에서 발생하는 소음의 주 원인으로 인식되고 있었다. 하지만, 압축기의 작동 속도가 증가함에 따라 압축기에 연결된 배관계에서의 냉매 유동 기인 진동에 의한 소음의 상대적 기여도가 증가하였다. 본 논문에서는 에어컨 압축기 배기 배관계에서의 유체 기인 소음을 수치적으로 예측할 수 있는 해석방법을 정립하였다. 이 단계에서, 해석 결과와 실험결과의 비교를 통해 해석의 신뢰성을 확인하였다. 추가적으로, 압축기 배기 배관계 방사 소음에 대하여 압축기 맥동음과 압축기 진동에 의한 소음의 영향을 주파수 대역별로 비교하였다. 압축기 진동에 의한 소음이 저주파수 대역에 기여함을 확인하였으며, 압축기 맥동음이 고주파수에서의 소음에 영향을 줌을 확인하였다.

• 대한기계학회논문집A
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• 제33권10호
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• pp.1108-1118
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• 2009

A Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source of the order of $950^{\circ}C$ for nuclear hydrogen generation, which can produce hydrogen from water or natural gas. A primary hot gas duct (HGD) as a coaxial double-tube type cross vessel is a key component connecting a reactor pressure vessel and an intermediate heat exchanger in the VHTR. In this study, a structural sizing methodology for the primary HGD of the VHTR is suggested in order to modulate a flow-induced vibration (FIV). And as an example, a structural sizing of the horizontal HGD with a coaxial double-tube structure was carried out using the suggested method. These activities include a decision of the geometric dimensions, a selection of the material, and an evaluation of the strength of the coaxial double-tube type cross vessel components. Also in order to compare the FIV characteristics of the proposed design cases, a fluid-structure interaction (FSI) analysis was carried out using the ADINA code.

• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1712-1724
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• 2024

Spiral tubes are used in a wide range of applications and it is significant to understand the vibration introduced by two-phase flow in spiral tubes. In this paper, the numerical method is used to study the vibration induced by the gas-liquid two-phase flow in a spiral tube with different flow regimes. The pressure fluctuation characteristics at the pipe wall and the solid vibration response characteristics are obtained. The results show that the motion of small bubbles in bubbly flow leads to small pressure fluctuations with low-frequency broadband (0-50 Hz). The motion of the gas plug in the plug flow causes small amplitude periodic pressure fluctuation with a shortened low-frequency broadband (0-15 Hz) compared to the bubbly flow. The motion of the gas slug in the slug flow causes large periodic fluctuations in pressure with a significant dominant frequency (6-7 Hz). The wavy flow is very stable and has a distinct main frequency (1-2 Hz). The vibration regime in the bubbly flow and wave flow are close to the first-order mode, and the vertical vibrating component is dominant. The plug flow and slug flow excite higher-order vibration modes, and the lateral vibration component plays more important part in the vibration response.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4561-4569
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• 2023

A correct understanding of vibration-based degradation is crucial from the standpoint of maintenance for Steam Generators (SG) as crucial mechanical equipment in nuclear power plants. This study has established a novel approach to developing a model for investigating tube bundle degradation according to crack growth caused by two-phase Flow-Induced Vibration (FIV). An important step in the approach is to calculate the two-phase flow field parameters between the SG tube bundles in various zones using the porous media model to determine the velocity and vapor volume fraction. Afterward, to determine the vibration properties of the tube bundles, the Fluid-Solid Interaction (FSI) analysis is performed in eighteen thermal-hydraulic zones. Tube bundle degradation based on crack growth using the sixteen most probable initial cracks and within each SG thermal-hydraulic zone is performed to calculate useful lifetime. Large Eddy Simulation (LES) model, Paris law, and Wiener process model are considered to model the turbulent crossflow around the tube bundles, simulation of elliptical crack growth due to the vibration characteristics, and estimation of SG tube bundles degradation, respectively. The analysis shows that the tube deforms most noticeably in the zone with the highest velocity. As a result, cracks propagate more quickly in the tube with a higher height. In all simulations based on different initial crack sizes, it was observed that zone 16 experiences the greatest deformation and, subsequently, the fastest degradation, with a velocity and vapor volume fraction of 0.5 m/s and 0.4, respectively.

• 한국소음진동공학회논문집
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• 제26권2호
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• pp.203-209
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• 2016

According to the wear detection history for the steam generator tubes in the nuclear power plant, the outer tubes inside the steam generator have more problems on the flow-induced vibration than inner tubes. Many researchers and engineers have used a specified added mass coefficient for a given tube array during the design stage of the steam generator even though the coefficient is not constant for entire tube in cylindrical shell. The aim of this study is to find out the distribution of added mass coefficients for each tube along the radial location. When numbers of tubes inside a cylindrical shell are increased, values of added mass coefficients are also increased. Added mass coefficients at outer tubes are less than those of inner tubes and they are decreased with increasing the gap between the outermost tube and the cylindrical shell. It also turns out when the gap between the outermost tube and the cylindrical shell approaches infinite value, the added mass coefficient converges to an asymptotic value of given tube array in a free fluid field.

• Tribology and Lubricants
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• 제24권6호
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• pp.362-367
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• 2008

Tubes in nuclear steam generators are held up by supports because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube-support. The fretting wear of tube-support can threaten the safety of nuclear power plant. The tube and support materials were Inconel 690 and STS 409. The wear tests were conducted in various environments, which are in water without flow, in flowing water and in flowing water with air. The results showed that the flow of water influenced on the wear-life of tube. The wear-life of tube decreased in water flow as compared with wear-life in stationary water.

• Tribology and Lubricants
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• 제23권3호
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• pp.130-133
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• 2007

Fuel cladding tubes in nuclear fuel assembly are held up by supporting grids because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube and support. The fretting wear of tube and support can threaten the safety of nuclear power plant. Therefore, a research about the fretting wear characteristics of tube-support is required. The fretting wear tests were performed with supporting grids and cladding tubes, especially after corrosion treatment on tubes, in water. The tests were done using various applied loads with fixed amplitude. From the results of fretting tests, the wear amounts of tube materials can be predictable by obtaining the wear coefficient using the work rate model. Due to stick phenomena the wear depth was changed as increasing load and temperature. The maximum wear depth was decreased as increasing the water temperatures. At high temperatures there are the regions of some severe adhesion due to stick phenomena.