• Nuclear Engineering and Technology
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• 제38권1호
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• pp.69-80
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• 2006

Fluid-elastic instability in an air-water two-phase cross-flow has been experimentally investigated using two different arrays of straight tube bundles: normal square (NS) array and rotated square (RS) array tube bundles with the same pitch-to-diameter ratio of 1.633. Experiments have been performed over wide ranges of mass flux and void fraction. The quantitative tube vibration displacement was measured using a pair of strain gages and the detailed orbit of the tube motion was analyzed from high-speed video recordings. The present study provides the flow pattern, detailed tube vibration response, damping ratio, hydrodynamic mass, and the fluid-elastic instability for each tube bundle. Tube vibration characteristics of the RS array tube bundle in the two-phase flow condition were quite different from those of the NS array tube bundle with respect to the vortex shedding induced vibration and the shape of the oval orbit of the tube motion at the fluid-elastic instability as well as the fluid-elastic instability constant.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1650-1657
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• 2019

Fluidelastic instability and nonlinear dynamics of tube bundles is a key issue in a steam generator. Especially, once the post-instability motion of the tube becomes larger than the clearance gap to other tubes, effective contact or impact between the tubes under consideration and the other tube inevitable. There is seldom theoretical analysis to the nonlinear dynamic characteristics of a tube array in two-phase flow. In this paper, experimental and numerical studies were utilized to obtain the critical velocity of the flow-induced instability of a rotated triangular tube array. The calculation results agreed well with the experimental data. To explore the post-instability dynamics of the tube array system, a Runge-Kutta scheme was used to solve the nonlinear governing equations of tube motion. The numerical results indicated that, when the flow pitch velocity is larger than the critical velocity, the tube array system is undergoing a limit cycle motion, and the dynamic characteristics of the tube array are almost similar for different void fractions.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.2026-2033
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• 2019

Fluidelastic instability of a tube array is a key factor of the security of a nuclear power plant. An unsteady model of the fluidelastic instability of a tube array subjected to two-phase flow was developed to analyze the fluidelastic instability of tube bundles in two-phase flow. Based on this model, a computational program was written to calculate the eigenvalue and the critical velocity of the fluidelastic instability. The unsteady model and the program were verified by comparing with the experimental results reported previously. The influences of void fraction and the tube's material properties on the critical velocity were investigated. Numerical results showed that, with increasing the void fraction of the two-phase flow, the tube array becomes more stable. The results indicate that the critical velocities of the tube array made of stainless are much higher than those of the other two tube arrays within void fraction ranging from 20% to 80%.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.530-537
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• 2020

This study investigates the effect of an included angle and heat flux on heat transfer of V-shape tube array having a horizontal upper tube. The test uses two stainless steel tubes with a smooth surface submerged under the water at atmospheric pressure. The angle varies from 2° to 24°. The heat transfer coefficient gets decreasing in consequence as the angle increases. The enhancement due to the lower tube is distinct as the heat flux is lower than 60 kW/㎡, where the effect of the convective flow is dominant. The present study and the published results show a similar tendency. Although the heat transfer coefficient for the present study is smaller than the symmetry case, enhanced heat transfer is observed compared to the tube array having a lower horizontal tube as the included angle is less than 10°.

• 한국압력기기공학회 논문집
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• 제5권1호
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• pp.25-31
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• 2009

Steam Generator(SG) tube is an important component of Nuclear Power Plant(NPP), which comprises of the pressure boundary of primary system. The integrity of SG tube has been confirmed by the eddy current test every outage. In Korea, Bobbin probe and MRPC probe have been generally used for the eddy current test. Meanwhile the usage of Array probe has gradually increased in U.S., Japan and other countries. In this study, we investigated the defect detection capability of the Array probe through its preliminary application to SG tube inspection. The Array probe has the equivalent capability in the defect detection and sizing as the conventional methods. Thus it is desirable that the Array probe is generally applied to SG tube inspection in the domestic NPPs.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.1118-1124
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• 2023

Experimental study on the fluidelastic instability (FEI) of a curved tube bundle in single phase downward cross flow is investigated for the design qualification and analysis input preparation of helical coiled steam generator tubing. A 6×9 normal square curved tube array with equal and different vertical/horizontal pitch-to-diameter ratio was under-tested up to 6 m/s in term of gap flow velocity to measure the critical velocity for FEI. The critical velocity for FEI was measured at the turning point from the vibration amplitude plot along the gap flow velocity. Our test results were compared with straight tube results and published data in the design guideline. The applicability of the current design guidelines to a curved tube bundle is also assessed. We found that introducing frequency difference in a curved tube array increases the critical velocity for fluidelastic instability.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.310-316
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• 2019

Fluidelastic instability is a key issue in steam generator tube bundles subjected in cross-flow. With a low flow velocity, a large amplitude vibration of the tube observed by many researchers. However, the mechanism of this vibration is seldom analyzed. In this paper, the mechanism of cross-flow effects on fluidelastic instability of tube bundles was investigated. Analysis reveals that when the system reaches the critical state, there would be two forms, with two critical velocities, and thus two expressions for the critical velocities were obtained. Fluidelastic instability experiment and numerical analysis were conducted to obtain the critical velocity. And, if system damping is small, with increases of the flow velocity, the stability behavior of tube array changes. At a certain flow velocity, the stability of tube array reaches the first critical state, a dynamic bifurcation occurs. The tube array returns to a stable state with continues to increase the flow velocity. At another certain flow velocity, the stability of tube array reaches the second critical state, another dynamic bifurcation occurs. However, if system damping is big, there is only one critical state with increases the flow velocity. Compared the results of experiments to numerical analysis, it shows a good agreement.

• 한국산학기술학회논문지
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• 제17권2호
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• pp.712-718
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• 2016

배열회수 보일러의 전열관군은 외부에 가스터빈에서 나온 고온의 배기가스가 흐르게 된다. 이러한 유체의 흐름으로 인해 전열관군에서 시간변화에 따라 양력과 항력의 변동이 발생하는데 이에 따라 유동 유발 진동이 발생한다. 이러한 진동이 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 유동 유발 진동에 대한 무차원 PSD(Power Spectral Density) 함수를 무차원 주파수인 Strouhal 수, fD/U의 함수로 실험적 결과들이 도출되어 있다. 본 연구는 열교환기 전열관군에서 유동 유발 진동에 관한 기존의 실험적 연구들의 결과를 전산유체해석을 통해 검증하고 배열회수 보일러의 전열관군의 양력과 항력의 PSD 특성에 적용하기 위한 기반을 마련하는 것을 목적으로 한다. 이러한 것을 위해 기존 연구에서 실험에 사용한 전열관군에서 비정상 상태 유동해석을 수행하여 전열관군에서의 양력과 항력 변화 특성을 살펴보았다. 또한 전열관군에서 양력과 항력 변동 특성으로부터 유동 유발 진동에 따른 PSD 특성 결과를 도출하여 기존의 연구들과 비교를 통해 전열관군에서의 PSD 특성을 정립하였다.

• 한국산학기술학회논문지
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• 제16권10호
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• pp.6641-6646
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• 2015

배열회수 보일러의 전열관군은 외부에 가스터빈에서 나온 고온의 배기가스가 흐르게 된다. 이러한 유체의 흐름으로 인해 전열관군에서 시간변화에 따라 양력의 변동이 발생하는데 이에 따라 유동 유발 진동이 발생한다. 이러한 진동이 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 유동 유발 진동에 대한 무차원 PSD(Power Spectral Density) 함수를 무차원 주파수인 Strouhal 수, fU/U의 함수로 실험적 결과들이 도출되어 있다. 본 연구는 열교환기 전열관군에서 유동 유발 진동에 관한 기존의 실험적 연구들의 결과를 전산유체해석을 통해 검증하고 배열회수 보일러의 전열관군의 유동 유발 진동 특성에 적용하기 위한 기반을 마련하는 것을 목적으로 한다. 이러한 것을 위해 기존 연구에서 실험에 사용한 전열관군에서 비정상 상태 유동해석을 수행하여 전열관군에서의 양력 변화 특성을 살펴보았다. 또한 전열관군에서 양력 변동 특성으로부터 유동 유발 진동에 따른 PSD 특성 결과를 도출하여 기존의 연구들과 비교를 통해 전열관군에서의 PSD 특성을 정립하였다.

• 설비공학논문집
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• 제19권3호
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• pp.245-252
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• 2007

The important problems from the point of view of preventing global warming are to save the power consumption of automotive air-conditioning systems and reduce the refrigerant amount filled. To achieve such requirements, integral receiver/dryer (R/D) condensers were developed recently. Typical integral R/D condensers have extra headers that play the role of R/D. Except an extra header and somewhat complex tube array resulting from the extra header, the most integral R/D condensers have almost the same specification that tube has multi channels, fin has louvers, flow in tube is parallel, etc. When integral condensers are applied, it is known that the refrigerating effect increases, resulting from the increase of subcooling degree in condenser, and the refrigerant amount used saves. In spite of several merits, integral condensers have not been applied a lot. That is why there is an uncertainty in performance, using integral condensers. The objective of this study is to theoretically optimize the tube array in an integral R/D condenser that is really being applied to some vehicles. The tube array has a great effect on the performance of the integral condenser as well as common ones. Through computer simulation, we could see that the tube array, 14-6-3-5-3-4, in the same condenser was the best, comparing heat release rate, pressure drop, etc. to the real array, 17-5-3-3-2-5. It should be noted that the optimization is based on the condenser performance only.