• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2005년도 춘계학술대회논문집
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• pp.312-317
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• 2005

Fluid-elastic instability and turbulence excitation for an under developing steam generator are investigated numerically. The stability ratio and the amplitude of turbulence excitation are obtained by using the PIAT (Program for Integrity Assessment of Steam Generator Tube) code from the information on the thermal-hydraulic data of the steam generator. The aspect ratio, the ratio between the height of U-tube from the upper most tube support plate (h) and the width of two vertical portion of U-tube (w), is defined for geometric parameter study. Several aspect ratios with relocation of tube support plates are adopted to study the effects on the mode shapes and characteristics of flow-induced vibration. When the aspect ratio exceeds value of 1, most of the mode shapes at low frequency are generated at the top of U-tube. It makes very high value of the stability ratio and the amplitude of turbulent excitation as well. We can consider that the local mode shape at the upper side of U-tube will develop the wear phenomena between the tube and the anti-vibration bars such as vertical, horizontal, and diagonal strips. It turns out that the aspect ratio reveals very important parameter for the design stage of the steam generator. The appropriate value of the aspect ratio should be specified and applied.

• 한국소음진동공학회논문집
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• 제15권9호
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• pp.1009-1015
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• 2005

Fluid-elastic instability and turbulence excitation for an under developing steam generator are investigated numerically. The stability ratio and the amplitude of turbulence excitation are obtained by using the $PIAT^{(R)}$ (program for integrity assessment of steam generator tube) code from the information on the thermal-hydraulic data of the steam generator. The aspect ratio, the ratio between the height of U-tube from the upper most tube support Plate (h) and the width of two vertical portion of U-tube (w), is defined for geometric parameter study. Several aspect ratios with relocation of tube support plates are adopted to study the effects on the mode shapes and characteristics of flow-induced vibration. When the aspect ratio exceeds value of 1, most of the mode shapes at low frequency are generated at the top of U-tube. It makes very high value of the stability ratio and the amplitude of turbulent excitation as well. We can consider that the local mode shape at the upper side of U-tube will develop the wear phenomena between the tube and the anti-nitration bars such as vortical, horizontal, and diagonal strips. It turns out that the aspect ratio reveals very important parameter for the design stage of the steam generator. The appropriate value of the aspect ratio should be specified and applied.

• 한국압력기기공학회 논문집
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• 제11권1호
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• pp.20-24
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• 2015

Steam generator(S/G) tubes in pressurized water reactor (PWR's) are subject to several types of degradation. This degradation includes denting, pitting, intergranular attack(IGA), intergranular stress corrosion cracking(IGSCC), fatigue, fretting and wear. Degradation can be derived from either the primary side(inside) or the secondary side(outside) of the tube. Recent issue for tube degradation in domestic steam generator is the tube end cracking on seal weld region. The seal weld region at the tube end and tube itself is regarded as a pressure boundary between the primary side and the secondary side. One of the Westinghouse Model-F S/G has experienced tube end cracking and its number of plugging approximately becomes to the operating limit up to 5% due to tube end cracking which was reported as SAI/MAI(single/multiple axial indication) or SCI/MCI(Single/multiple circumferential indication) from the results of eddy current testing. Eddy current mock-up test was carried out to determine the origin of cracking whether it is from weld zone area or parent tube. This result was helpful to analyze crack location on ECT data. Correct action on this problem was the installation of tube-end sleeve. Last year, after removing 340 installed plugs from tubes, selected 269 tubes took tube-end sleeve installation. Tube-end sleeve brought pressure boundary from parent tube to installed sleeve tube. Tube-end sleeve has the benefit of reducing outage period and increasing more revenue than replacing S/G. This paper is provided to assist interest parties in effectively understanding this issue.

• 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.

• 대한기계학회논문집A
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• 제28권7호
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• pp.1006-1013
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• 2004

Operating experience of steam generators shows that the tubes are degraded by stress corrosion cracking, fretting wear and so on. These defected tubes could stay in service if it is proved that the tubes have sufficient structural margin to preclude the risk of tube bursting. This paper provides detailed plastic limit pressure solutions for through-wall cracks in the steam generator tubes. These are developed based on three dimensional(3D) finite element analyses assuming elastic-perfectly plastic material behavior. Both axial and circumferential through-wall cracks in free span and in u-bend regions are considered. The resulting limit pressure solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.

• 한국전산유체공학회지
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• 제9권2호
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• pp.11-17
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• 2004

CFD analysis was performed to figure out flow behavior in the inlet plenum of new research reactor where coolant is injected to the flow tubes with the fuel assembly. The computation results showed that large-scale vortices are generated in the inlet plenum by flow stream injected from inlet pipe. These vortices are divided into small vortices and reversed their revolution. They may lead to flow-induced vibration of fuel assembly, moreover, which has been regarded as a cause of fretting wear of fuel assembly. Also there is an another important thing that average velocity of each flow-tube is uneven showing difference in maximum 18%. So it has been suggested that perforated baffle will be installed to prevent the formation of vortex in the inlet plenum. Two perforated baffles, one is flow skirt and the other is muffler type flow straightener, were proposed and their effect was evaluated using commercial CFD code, Fluent. According to CFD analysis for two perforated baffles, it was confirmed that both of them can prevent or reduce vortex formation in the inlet plenum and make average velocity of each flow tube more even.

• Journal of Mechanical Science and Technology
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• 제18권1호
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• pp.74-81
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• 2004

The KSNP Steam Generators (Youngkwang Unit 3 and 4, Ulchin Unit 3 and 4) have a problem of U-tube fretting wear due to Flow Induced Vibration (FIV). In particular, the wear is localized and concentrated in a small area of upper part of U-bend in the Central Cavity region. The region has some conditions susceptible to the FIV, which are high flow velocity, high void fraction, and long unsupported span. Even though the FIV could be occurred by many mechanisms, the main mechanism would be fluid-elastic instability, or turbulent excitation. To remedy the problem, Eggcrate Flow Distribution Plate (EFDP) was installed in the Central Cavity region or Ulchin Unit 5 and 6 steam generators, so that it reduces the flow velocity in the region to a certain level. However, the cause of the FIV and the effectiveness of the EFDP was not thoroughly studied and checked. In this study, therefore the Stability Ratio (SR), which is the ratio of the actual velocity to the critical velocity, was compared between the value before the installation of EFDP and that after. Also the possibility of fluid-elastic instability of KSNP steam generator and the effectiveness of EFDP were checked based on the ATHOS3 code calculation and the Pettigrew's experimental results. The calculated results were plotted in a fluid-elastic instability criteria-diagram (Pettigrew, 1998, Fig. 9). The plotted result showed that KSNP steam generator with EFDP had the margin of Fluid-Elastic Instability by almost 25％.

• Nuclear Engineering and Technology
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• 제46권5호
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• pp.633-640
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• 2014

In this research, the surface roughness affecting the pressure drop in a pipe used as the steam generator of a PWR was studied. Based on the CFD (Computational Fluid Dynamics) technique using a commercial code named ANSYS-FLUENT, a straight pipe was modeled to obtain the Darcy frictional coefficient, changed with a range of various surface roughness ratios as well as Reynolds numbers. The result is validated by the comparison with a Moody chart to set the appropriate size of grids at the wall for the correct consideration of surface roughness. The pressure drop in a full-scale U-shaped pipe is measured with the same code, correlated with the surface roughness ratio. In the next stage, we studied a reduced scale model of a U-shaped heat pipe with experiment and analysis of the investigation into fluid-structure interaction (FSI). The material of the pipe was cut from the real heat pipe of a material named Inconel 690 alloy, now used in steam generators. The accelerations at the fixed stations on the outer surface of the pipe model are measured in the series of time history, and Fourier transformed to the frequency domain. The natural frequency of three leading modes were traced from the FFT data, and compared with the result of a numerical analysis for unsteady, incompressible flow. The corresponding mode shapes and maximum displacement are obtained numerically from the FSI simulation with the coupling of the commercial codes, ANSYS-FLUENT and TRANSIENT_STRUCTURAL. The primary frequencies for the model system consist of three parts: structural vibration, BPF(blade pass frequency) of pump, and fluid-structure interaction.

• 비파괴검사학회지
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• 제32권1호
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• pp.64-70
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• 2012

국내 원자력발전소 증기발생기 전열관의 건전성을 평가하기 위해서 수행되는 와전류검사의 탐촉자는 주로 두가지 종류가 사용한다. 첫 번째 와전류탐촉자는 마모와 같은 체적성 결함을 검사하기 위해 사용되는 보빈탐촉자이다. 두 번째 와전류탐촉자는 균열과 같은 비체적성 결함을 검사하기 위한 회전형 탐촉자이다. 와전류탐촉자는 검사 계통의 핵심적인 부분으로서 특정 절차서에 따라 평가가 이루어질 때 대상 시험체의 합부를 결정하는 자료를 제공하게 된다. 또한, 수집된 와전류신호의 품질은 사용되는 탐촉자의 설계특성, 기하학적 형태, 운전주파수에 따라 결정되고, 검사 결과에 미치는 영향이 크기 때문에 와전류검사 탐촉자의 선정은 특히 중요하다. 본 연구에서는 차동형 보빈탐촉자의 코일 권선 수의 변화가 탐촉자의 전기적 특성에 미치는 영향을 분석하였다. 이 결과를 이용하여 원전 증기발생기 전열관 와전류검사 보빈탐촉자를 설계하였다. 연구 결과 코일 권선 수의 변화는 전열관 형상 및 재질에 따른 검사주파수 선정에 크게 영향이 미침을 알 수 있다. 따라서 본 연구 결과를 통하여 원전 증기발생기 전열관 와전류검사 보빈탐촉자의 설계시 더욱 정밀한 코일 권선 수 설정에 토대를 구축하였다.