• 대한기계학회논문집
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• 제17권1호
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• pp.200-214
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• 1993

본 연구에서는 Westinghouse Model 51 증기발생기의 U-bend 영역에서 2차측 유체의 횡단유동으로 유발될 수 있는 튜브군의 유체탄성불안정을 예측하기 위한 해석 을 수행하고 그 대표적인 결과들을 제시하였다. 그리고 U-bend 영역에서 AVB에 의한 튜브의 지지상태와 형태 및 최상부 TSP에서 Denting 또는 이물질 고착으로 인하여 변 경된 튜브의 고정지지조건 등이 튜브의 유체탄성불안정 응답에 미치는 영향을 조사하 였다. 유체탄성불안정 해석과정에서 필수적으로 선행되어야 하는 2차측 3차원 2상 유동장 계산은 증기발생기 열수력 해석용인 ATHOS3 코드로써 수행되었으며, U-튜브의 고유진동수와 모우드 형상은 공학해석용 유한요소 프로그램인 ANSYS코드로써 계산되었 다.

• 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.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2006년도 추계학술대회 논문집
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• pp.361-362
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• 2006

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.828-833
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• 2001

In this study, an analysis tool to assess the susceptibility of steam generator tubes due to the flow-induced vibration was developed. The fluid-elastic instability analysis of the U-tube bundle for CE-type steam generator was accomplished. The effective mass distribution along the U-tube was obtained to calculate the natural frequency and dynamic mode shape. Finally, stability ratios for selected tubes are obtained.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2001년도 추계학술대회논문집 II
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• pp.1400-1405
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• 2001

In this study, an analysis program to assess the susceptibility of steam generator tubes due to the flow-induced vibration was developed. Analysis of fluid-elastic instability and random turbulence excitation for the U-tube bundle in CE-type steam generator was accomplished. The effective mass distribution along the U-tube was obtained to calculate the natural frequency and dynamic mode shape. Finally, stability ratios and rms vibration amplitude for selected tubes are obtained.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.213-218
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• 1996

A simulation model for the transient behavior of CANDU U-tube steam generator(UTSG) has been developed for application to the simulation of operational transient behavior of CANDU nuclear power plant. For application to CANDU UTSG. tile design characteristics of CANDU UTSG such as Wolsong Units, feedwater inlet near the tube sheet. is approximated. For realistic prediction of thermal hydraulic behavior of and tube bundle region is divided into two separate control volumes, subcooled region and saturated region. and the variation of thermal hydraulic properties within a control volume is considered. Test results for typical CANDU operational transient case show reasonable transient behavior of steam generator and considered to be applicable to the simulation of overall plant.

• Nuclear Engineering and Technology
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• 제42권1호
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• pp.97-108
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• 2010

Fluid-elastic instability and turbulence-induced vibration of steam generator U-tubes of a nuclear power plant are studied numerically to investigate the effect of design changes of support structures in the upper region of the tubes. Two steam generator models, Model A and Model B, are considered in this study. The main design features of both models are identical except for the conditions of vertical and horizontal support bars. The location and number of vertical and horizontal support bars at the middle of the U-bend region in Model A differs from that of Model B. The stability ratio and the amplitude of turbulence-induced vibration are calculated by a computer program based on the ASME code. The mode shape with a large modal displacement at the upper region of the U-tube is the key parameter related to the fretting wear between the tube and its support structures, such as vertical, horizontal, and diagonal support bars. Therefore, the location and the number of vertical and horizontal support bars have a great influence on the fretting wear mechanism. The variation in the stability ratios for each vibrational mode is compared with respect to Model A and Model B. Even though both models satisfy the design criteria, Model A shows substantial improvements over Model B, particularly in terms of having greater amplitude margins in the turbulence-excited vibration (especially at the inner region of the tube bundle) and better stability ratios for the fluid-elastic instability.

• 대한기계학회논문집A
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• 제31권7호
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• pp.784-791
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• 2007

Fretting, which is a special type of wear, is defined as small amplitude tangential oscillation along the contacting interface between two materials. In nuclear power plants, fretting wear caused by flow induced vibration (FIV) can make a serious problem in a U-tube bundle in steam generator. In this study, substructure method is developed and is verified the feasibility for the finite element model of fretting wear problems. This method is applied to the two-dimensional finite element analyses, which simulate the contact behavior of actual tube to support. For these examples, computing time can be reduced up to 1/5 in comparisons with conventional finite element analyses.