• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집B
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• pp.105-111
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• 2000

A numerical study has been performed to obtain the heat transfer and pressure drop characteristics for shell-and-tube heat exchanger with various shapes of tubes. The Tubes have variation of Aspect Ratio, Pitch and Rotation. Results are presented as plots of Colburn j factor and friction factor f against Aspect Ratio, Pitch and Rotation. As Aspect Ratio increases, j factor and f factor decreases. As Pitch increases, j factor decreases. j/f have optimized Pitch for each Aspect Ratio. Accordingly, there is fitness of Aspect Ratio and Pitch fur most effective cases. The Rotation of tubes are of no meaning for both heat transfer and pressure drop.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2007년도 춘계 학술대회논문집
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• pp.149-152
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• 2007

The numerical simulations on the heat transfer and flow field were carried out for the improvement of the performance of the shell and tube heat exchanger. The steady incompressible 3-D Navier-Stokes solution is obtained with the actual operational condition and geometry of the heat exchanger. The present geometry of the heat exchanger causes poor heat transfer since the air inside shell dose not flow through the tube bundle, but around it. The enhancement of the heat transfer can be achieved by the variation of the design factor like the sealing strip located on the top/bottom and middle of the baffle.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(4)
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• pp.312-318
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• 1996

Flow induced vibration experiments were done to determine the vibration magnitude of tubes in the economizer tube lane region of a steam generator. The objective of this experiment is to demonstrate that the tubes are not susceptible to harmful levels of vibration at 100% of full power flow and to quantify the remaining design margin at 120% and 140% of full power flow.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.534-539
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• 1997

A CHF table method has been assessed in this study for rod bundle CHF predictions. At the conceptual design stage for a new reactor, a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. In many aspects, a CHF table method (i.e., the use of a round tube CHF table with appropriate bundle correction factors) can be a promising way to fulfill this need. So the assessment of the CHF table method has been performed with the bundle CHF data relevant to pressurized water reactors (PWRs). For comparison purposes, W-3R and EPRI-1 were also applied to the same data base. Data analysis has been conducted with the subchannel code COBRA-IV-I. The CHF table method shows the best predictions based on the direct substitution method. Improvements of the bundle correction factors, especially for the spacer grid and cold wall effects, are desirable for better predictions. Though the present assessment is somewhat limited in both fuel geometries and operating conditions, the CHF table method clearly shows potential to be a general CHF predictor.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.901-910
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• 2021

KHNP recently has obtained the approval for the commercialization of the modified 37-element (or 37 M) fuel bundle and now is loading the 37 M fuel bundles in CANDU-6 reactors in KOREA. One of the main issues for approval was the burnup limit. Due to CANDU design and neutronic characteristics, there was no specific burnup restriction of a fuel bundle. The absence of a burnup limit does not mean that a fuel bundle can stay in the CANDU reactor without a time limit. However, the regulator requested traditional design values as well as the burnup limit reflecting the computer code uncertainty. The method for the PWR burnup limit was not applied to the CANDU fuel bundle. Since there was no approved methodology to build the burnup limit with uncertainties, KHNP introduced a Monte-Carlo method coupled with a 95/95 approach to determine the conservative burnup limit from the viewpoint of the centerline temperature, internal pressure, strain measurement deviation. Moreover, to consider the uncertainties of various computing models, a converted power uncertainty was introduced. This paper presents the methodology and puts forward the limits on burnup, evaluated for each of the existing and modified fuel bundles, in consideration of the pressure tube aging condition.

• 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.

• 한국소음진동공학회논문집
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• 제26권2호
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• pp.203-209
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• 2016

According to the wear detection history for the steam generator tubes in the nuclear power plant, the outer tubes inside the steam generator have more problems on the flow-induced vibration than inner tubes. Many researchers and engineers have used a specified added mass coefficient for a given tube array during the design stage of the steam generator even though the coefficient is not constant for entire tube in cylindrical shell. The aim of this study is to find out the distribution of added mass coefficients for each tube along the radial location. When numbers of tubes inside a cylindrical shell are increased, values of added mass coefficients are also increased. Added mass coefficients at outer tubes are less than those of inner tubes and they are decreased with increasing the gap between the outermost tube and the cylindrical shell. It also turns out when the gap between the outermost tube and the cylindrical shell approaches infinite value, the added mass coefficient converges to an asymptotic value of given tube array in a free fluid field.

• 한국산학기술학회논문지
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• 제17권6호
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• pp.18-25
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• 2016

대형 압축식 냉동기의 증발기 전열관으로는 그간 평활관이 주로 사용되어 왔으나 최근들어 비등 성능이 우수한 성형 가공관을 많이 사용하고 있다. 증발기는 관군으로 구성되고 따라서 증발기 내 위치에 따라 냉매 상태가 다르다. 특히 R-11, R-123과 같은 저압 냉매는 압력변화에 따른 포화온도 변화가 크므로 위치에 따라 포화온도가 다르게 된다. 따라서 증발기를 적절히 설계하려면 증발기 내 각 위치에서의 냉매의 상태를 적절히 예측하여야 한다. 본 연구에서는 대형 냉동기의 만액식 증발기를 모사할 수 있는 컴퓨터 프로그램을 개발하였다. 이 프로그램은 증발기를 미소 체적으로 구분하고 각 미소 체적에 적절한 관 내외측 열전달 및 압력손실 상관식을 적용하여 해석을 수행하였다. 본 프로그램을 R-123을 사용하는 T사의 만액식 증발기 해석에 적용한 결과 만족할 만한 결과를 얻었다. 이 프로그램을 이용하여 신 냉매인 R-123, R-134a를 사용하는 만액식 증발기의 해석을 수행하였고 특히 관군 세장비의 영향을 검토하였다.

• 대한기계학회논문집B
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• 제41권3호
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• pp.171-181
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• 2017

2상 횡 유동은 응축기, 증발기와 원자력의 증기 발생기와 같은 열교환기의 튜브와 셀 사이에 존재한다. 공기/물의 2상 유동에 놓인 관군에 작용하는 항력을 실험적으로 평가하였다. 2상 유동에 놓인 관군은 정사각형 배열이다. 피치 직경 비는 1.35이었고, 실린더의 직경은 18 mm이다. 관군에 유동방향으로 작용하는 항력을 측정하여 항력계수와 2상 유동 감쇠비를 계산하였다. 2상 유동 감쇠비는 균질 2상 유동의 이론식을 사용하여 구하여 실험의 결과와 비교하였다. 압력과 항력의 상관계수를 실험결과를 고려하여 평가하였다. 상관계수는 이론적으로 항력을 계산할 때에 사용된다. 질량유량을 증가할수록 측정된 항력으로부터 구한 항력계수와 감쇠비가 균질 유동의 이론적 결과와 잘 일치함을 보이고 있다. 결과적으로 충분히 큰 질량 유량의 기포 유동인 경우에는 감쇠비를 균질 유동에 근거한 이론식으로 계산할 수 있다.

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