• 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.

• Journal of Advanced Marine Engineering and Technology
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• 제24권3호
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• pp.113-120
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• 2000

On the absorber of absorption chiller/heater, LiBr solution at high concentration is sprinkled on a bundle of horizontal tube cooled by cooling water. In this case, the conditions of LiBr solution and cooling water have an influence on heat/mass transfer coefficient in this system. Therefor it is important to find optimal operation conditions of absorption chiller/heater to save energy. Heat and mass transfer coefficient increased with the increase of solution flow rate, and also heat and mass transfer rate increased but overall heat and mass transfer coefficient decreased by increasing the solution concentration within the experimental range. The superheating of the solution resulted in superior heat transfer character to a state of equilibrium from the point of heat flux and overall heat transfer coefficient.

• 설비공학논문집
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• 제11권3호
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• pp.332-338
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• 1999

An experimental study of absorption phenomena of water vapor into LiBr solution was carried out to find out the optimum solution flow rate. The staggered bundle of horizontal absorption tubes, which are the same configuration as the commercial heat pump, were tested. The results showed that the heat transfer and absorption rate were enhanced with the increase of LiBr solution flow rate. Those values for different absorber pressures showed the same qualitative trends. The optimum flow rate of solution was obtained as three times of the designed flow rate.

• 설비공학논문집
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• 제11권5호
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• pp.570-578
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• 1999

An experimental study has been performed on heat and mass transfer in a falling film absorber with a strong lithium bromide solution in small-sized household absorption chiller/heater. Components were concentrically arranged in a cylindrical form. from the center, low temperature generator, absorber and evaporator. This arrangement of helical-typed heat exchangers allows to make the machine much more compact than conventional one. Experimental measurements were conducted with a helical absorber and the obtained data were compared with data in the literatures. The comparison revealed that the helical absorber tube provides a similar performance to existing tube bundle absorber in heat and mass transfer. As a result, the heat and mass transfer characteristics of helical type absorber showed the possibility of the reduction in size and weight of small] capacity absorption chiller/heater.

• 한국압력기기공학회 논문집
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• 제4권2호
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• pp.34-39
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• 2008

Steam Generator(SG) tubes are classified as one of the key components in nuclear power plants, and they should be periodically examined by the intensified management program for the assurance and diagnosis of their structural integrity. In this study, we use the optimum analysis technique to draw the detection and categorization of bowing(BOW) signals; abnormal tube-to-tube proximity in the SG upper bundle free span area. The locations in which BOW signals are detected likely have latent degradation of ODSCC(Outer Diameter Stress Corrosion Cracking). For the sake of timely and correct detection of BOW signals and diagnosis of ODSCC, we carried out the experimental demonstrations using a reduced mock-up. And we validated the MRPC(Motorized Rotating Pancake Coil) analysis technique is better than the bobbin. Hence, it comes to conclusion that the optimum analysis technique can be a good alternative for the reliable SG tube examination.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.2303-2308
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• 2007

Experiments were performed to determine the thermal (or turbulent) diffusion coefficient (TDC) and to investigate the critical heat flux (CHF) performance in the 5${\times}$5 rod bundle with 5 unheated rods which are supported by Hybrid Mixing Vane. In this study, HFC-134a fluid was used as working fluid and the fluid temperature were measured in the important subchannels. To determine the TDC value, the measured fluid temperatures were compared with the predicted values obtained from the MATRA code. The best optimized value of ${\beta}$ was found to be 0.02 by considering prediction statistics, i.e., average and standard deviations of the differences between the experimental results and code calculations. Using the best optimized value of ${\beta}$ as 0.02, the MATRA code predicts the test results of the fluid temperature within ${\pm}$1.0 % of error. According to the experimental results on CHF of 5 non-heating guide tubes, the case with non-heating guide tube showed a little good performance in terms of CHF.

• 대한기계학회논문집B
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• 제41권1호
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• pp.1-8
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• 2017

대기압 하에서 포화 상태를 유지하는 물의 내부에 잠긴 탠덤 튜브에 적용하기 위한 실험적 상관식을 새롭게 개발하였다. 상관식은 다발효과를 계산하기 위한 것으로 실험에서 측정한 값과 최소자승법을 사용하여 결정하였다. 상관식의 적절함을 평가하기 위해 통계분석을 수행하였다. 상관식은 실험값을 ${\pm}8%$ 범위 내에서 잘 예측함을 확인하였다. 상관식의 적용 범위는 피치=28.5~114mm, 방위각=$0^{\circ}{\sim}90^{\circ}$, 경사각=$0^{\circ}{\sim}90^{\circ}$, 상부 및 하부 튜브 열유속=$0{\sim}120kW/m^2$이다.

• 대한기계학회논문집A
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• 제26권11호
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• pp.2305-2311
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• 2002

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and heavy water coolant. Pressure tubes are installed horizontally inside the reactor and only selected samples are periodically examined during in-service inspection. In this respect, a probabilistic safety assessment method is more appropriate fur the assessment of overall pressure tube safety. The failure behavior of CANDU pressure tubes, however, is governed by delayed hydride cracking which is the major difference from pipings and reactor pressure vessels. Since the delayed hydride cracking has more widely distributed governing parameters, it is impossible to apply a general PFM methodology directly. In this paper, a PFM methodology for the safety assessment of CANDU pressure tubes is introduced by applying Monte Carlo simulation in determining failure probability Initial hydrogen concentration, flaw shape and depth, axial and radial crack growth rate and fracture toughness were considered as probabilistic variables. Parametric study has been done under the base of pressure tube dimension and hydride precipitation temperature in calculating failure probability. Unstable fracture and plastic collapse are used for the failure assessment. The estimated failure probability showed about three-order difference with changing dimensions of pressure tube.

• 대한기계학회논문집
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• 제15권6호
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• pp.1948-1966
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• 1991

본 연구에서는 2상 횡유동을 받는 튜브군의 진동 메카니즘을 규명하기 위한 실험계획의 일환으로 실시된 실험으로부터 튜브군의 유체탄성 불안정성 상수에 관해 고찰하였다. 실험은 먼저 p/d=1.47 및 1.32 튜브군에 대해 수행되었는데, 이들 튜브 군의 결과는 참고문헌에 발표하였다. 본 논문은 후속 실험으로 수행된 p/d=1.22인 튜브군을 사용하여 유체탄성 불안정성 상수를 고찰한 참고문헌의 후속논문이다. 실 험은 액체상태로 부터 99% 보이드율(void fraction)까지 변화된 2상 유동에서 튜브가 유체탄성 불안정성 상태에 도달할 때까지 점진적으로 증가하였다.실험결과는 p/d= 1.32 alc 1.47 튜브군의 유체탄성 불안정성 결과들과 종합. 비교되었다.

• 대한기계학회논문집A
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• 제24권1호
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• pp.206-214
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tube s. the integrity evaluation must be carried out. and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire: integrity evaluation process. For this reason. an integrity evaluation system, which provides efficient of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). This system does not only provide various databases including the 3-D finite element analysis results on pressure tubes, inspection data and design specifications but also is compatible with other commercial database software. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.