• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2011년도 춘계학술대회 논문집
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• pp.693-699
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• 2011

The outermost SG tubes have more structural problems than inside tubes. Many studies have used a uniform added mass coefficient for all of the SG tubes during the FIV analysis. The purpose of this study is to find out the added mass coefficients for each tube in a cylindrical shell. When a number of tubes are increased, added mass coefficients are also increased. And added mass coefficients at outermost tubes are less than those of inside tubes. According to gap changes between outermost tube and cylindrical shell, added mass coefficients are decreased with increasing the gap. When the gap has very large value, it shows that the added mass coefficient is asymptotically converged to the value of the tube array in a free fluid field.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2003년도 춘계학술대회논문집
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• pp.495-500
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• 2003

The vibration of a flexible cantilever tube with nonlinear constraints when it is subjected to flow internally with fluids is examined by experiment and theoretical analysis. These kind of studies have often been performed that finds the existence of chaotic motion. In this paper, the important parameters of the system leading to such a chaotic motion such as Young's modulus and coefficient of viscoelasticity in tube material are discussed. The parameters are investigated by means of a system identification so that comparisons are made between numerical analysis using the parameters of a handbook and the experimental results. The chaotic region led by several period-doubling bifurcations beyond the Hopf bifurcation is also re-established with phase portraits and bifurcation diagram so that one can define optimal parameters for system design.

• 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.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2002년도 춘계학술대회논문집
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• pp.883-888
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• 2002

This study describes the limit and application of the two-microphone impedance tube method to the sound transmission loss measurement of several sound isolation materials with different physical properties. For the sound isolation materials having small flexural rigidity, it is shown that the two-microphone impedance tube method is validated to practically measure the sound transmission loss. For the sound isolation materials having large flexural rigidity, on the other hand, it is found that the two-microphone impedance tube method is no longer valid to measure the sound transmission loss because the regions of resonance and mass law are moved into the higher frequencies. In addition, in order to accurately measure the sound transmission loss of sound isolation materials, their size should be decided based on the consideration of the effect of acoustic excitation on their vibration response.

• 한국가스학회지
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• 제20권2호
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• pp.23-29
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• 2016

배열회수 보일러의 전열관군은 외부에 고온의 배기가스가 흐르면서 유동 유발 진동을 야기 시키며 배열회수 보일러의 전열관군에서 파손을 야기할 수 있어서 열교환기의 구조적 안정성을 위해 열교환기의 전열관군에서 유동 유발 진동 특성을 규명할 필요가 있다. 일반적인 열교환기 전열관군에서 유동 유발 진동에 관한 실험적 연구는 기존에 많이 진행되어 오고 있으며 단일 원관이나 전열관군의 원관들에서 유동 유발 진동에 대한 무차원 PSD(Power Spectral Density) 함수를 무차원 주파수인 Strouhal 수, fD/U의 함수로 도출된 실험적 결과들이 도출되어 있다. 본 연구는 배열회수 보일러에 사용하는 휜튜브 전열관군에서 유동 유발 진동 특성을 규명하는 것을 목적으로 한다. 이러한 것을 위해 단일 휜튜브에서 비정상 상태 유동해석을 수행하여 주기적인 와동 발생 특성과 단일 휜튜브에서의 양력과 항력 변화 특성을 살펴보았다. 또한 단일 휜튜브에서 양력과 항력 변동 특성으로부터 유동 유발 진동에 따른 PSD 특성 결과를 도출하여 기존에 단순 원관에서 이루어졌던 연구들과 비교를 통해 단일 휜튜브 주위의 PSD 특성을 정립하였다.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2002년도 추계학술대회논문집
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• pp.927-932
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• 2002

The stability ratio and vibrational amplitude of each tube inside a steam generator have different values. We estimate the characteristics of flow-induced vibration for CE type steam generator with various column and row number of the tube. To obtain the thermal-hydraulic data and stability ratio we use the ATHOS3-MODI and PIAT-FEI/TE code respectively. It turns out that the steam generator has a bounded central zone with the distributed values of the stability ratio and the vibrational amplitude, and those values across the zone boundary become decreased.

• 대한기계학회논문집
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• 제10권1호
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• pp.1-6
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• 1986

본 논문에서는 제트유동에 놓인 튜우브 집합체의 진동 발생구조를 깊이 파악하고 일반성 있게 사용할 수 있는 설계 지침을 마련하기 위하여 보다 광법위 하게 수행된 파라미터 연구 내용을 다루었다

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2002년도 추계학술대회논문초록집
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• pp.390.2-390
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• 2002

The stability ratio and vibrational amplitude of each tube inside a steam generator have different values. We estimate the characteristics of flow-induced vibration for CE type steam generator with various column and row number of the tube. To obtain the thermal-hydraulic data and stability ratio we use the ATHOS3 and PIAT-FEI/TE code respectively. It turns out that the steam generator has a bounded central zone which the distributed values of the stability ratio and the vibrational amplitude, and those values across the zone boundary become decreased.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.821-830
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• 2009

In this paper, modal testing and finite element modeling results to identify the modal parameters of a nuclear fuel rod as well as its cladding tube are discussed. A vertically standing full-size cladding tube and a fuel rod with lead pellets were used in the modal testing. As excessive flow-induced vibration causes a failure in fuel rods, such as fretting wear, the vibration level of fuel rods should be low enough to prevent failure of these components. Because vibration amplitude can be estimated based on the modal parameters, the dynamic characteristics must be determined during the design process. Therefore, finite element models are developed based on the test results. The effect of a lumped mass attached to a cladding tube model was identified during the finite element model optimization process. Unlike a cladding tube model, the density of a fuel rod with pellets cannot be determined in a straightforward manner because pellets do not move in the same phase with the cladding tube motion. The density of a fuel rod with lead pellets was determined by comparing natural frequency ratio between the cladding tube and the rod. Thus, an improved fuel rod finite element model was developed based on the updated cladding tube model and an estimated fuel rod density considering the lead pellets. It is shown that the entire pellet mass does not contribute to the fuel rod dynamics; rather, they are only partially responsible for the fuel rod dynamic behavior.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.302-307
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• 2004

The major flow-induced vibration mechanisms such as fluid-elastic and turbulence excitation can cause the various types of wear of the steam generator tubes in unclear power plant. It is generally accepted that the tube wear due to vibration is affected by the presence of gap clearance between tube and support plate. Connors showed that the tube wear depth could be estimated by using the relationship between wear volume and sliding distance for contact time. Au-Yang predicted the wear depth by using the nonlinear characteristics of normal work rate to contact time. In this study the effect of gap size on the steam generator tubes wear is analyzed by deriving the wear depth versus normal work rate relationship from these previous results.