• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.1592-1597
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• 2004

Experimental studies were carried out to confirm the turbulent enhancement of the cooling system of nuclear reactor by large scale vortex generation in nuclear fuel rod bundle. The large scale vortex motions were generated by rearranging the inclination angles of mixing vanes to the coordinate directions. Experimental studies were carried out at Reynolds Number 60,000 with hydraulic condition. Normal variations of mean velocity and turbulent intensity in the rod bundle subchannel were measured by the 2-color LDV measurement system. The turbulence generated by split mixing vanes has small length scales so that they maintain only about 10DH after the spacer grid. On the other hand, the turbulences generated by the large scale vortex continue more and remain up $25D_{H}$ after the spacer grid.

• 대한기계학회논문집
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• 제10권1호
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• pp.7-14
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• 1986

본 연구에서는 가압경수형 원자로심을 모의하는 3*3배열로 된 모의연료 봉다발의 실험장치를 이용하여 재완수과정의 유동특성과 열전달특성을 파악하였으며, 재완수과정중 연료봉의 온도거동을 예측하는 REFLUX코드를 최근 개발된 연구자료를 토대로 수정하여 본실험결과와 비교하였다.

• 대한기계학회논문집
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• 제16권11호
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• pp.2111-2118
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• 1992

본 연구에서는 원관 내 및 평판 위의 흐름에 있어서 잘 확립된 벽면의 법칙을 적용하여 봉다발의 열전달계수를 예측하여 보고자 한다.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2479-2490
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• 2020

We performed experiments to measure a single-phase upward flow in a 5 × 5 rod bundle with spacer grids using a particle image velocimetry, focusing on the crossflow. The Reynolds number based on the hydraulic diameter and the bulk velocity is 10,000. The ratio of pitch between rods and rod diameter is 1.4 and spacer grid is installed periodically. The turbulence in the rod bundle results from the combination of a forced mixing and natural mixing. The forced mixing by the spacer grid persists up to 10D_h from the spacer grid, while the natural mixing is attributed to the crossflow between adjacent subchannels. The combined effects contribute to a sinusoidal distribution of the time-averaged stream-wise velocity along the lateral direction, which is relatively weak right behind the spacer grid as well as in the gap. The streamwise and lateral turbulence intensities are stronger right behind the spacer grid and in the gap. Based on these findings, we newly defined a modified mixing coefficient as the ratio of the lateral turbulence intensity to the time-averaged streamwise velocity, which shows a spatial variation. Finally, we compared the developed model with the measured data, which shows a good agreement with each other.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2005년도 창립60주년 기념 추계 학술대회
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• pp.97.1-97.1
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• 2005

• 한국전산유체공학회지
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• 제21권4호
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• pp.71-77
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• 2016

The CUPID code is a transient, three-dimensional, two-fluid, thermal-hydraulic code designed for a component-scale analysis of nuclear reactor components. The primary objective of this study is to assess the applicability of CUPID to single-phase turbulent flow analyses of $2{\times}2$ rod bundle subchannel. The bulk velocity at the inlet varies from 1.0 m/s up to 2.0 m/s which is equivalent to the fully turbulent flow with the range of Re=12,500 to 25,000. Adiabatic single-phase flow is assumed. The velocity profile at the exit region is quantitatively compared with both experimental measurement and commercial CFD tool. Three different boundary conditions are simulated and quantitatively compared each other. The calculation results of CUPID code shows a good agreement with the experimental data. It is concluded that the CUPID code has capability to reproduce the turbulent flow behavior for the $2{\times}2$ rod bundle geometry.

• 대한설비공학회:학술대회논문집
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• 대한설비공학회 2007년도 동계학술발표대회 논문집
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• pp.191-196
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• 2007

This paper presents the measuring technique for local heat transfer coefficients using a copper sensor in a rod bundle with mixing vanes. A copper sensor consists of a cartridge heater and four pieces of thermocouple. The Heater is located at the center of the copper sensor and thermocouples measure the surface temperature of the copper sensor. Unheated copper sensor and heated copper sensor are able to measure the local heat transfer coefficient at the position where the heated copper sensor is installed. However the entire region of a rod bundle is actually not heated, the decay of local heat transfer coefficients measured represents overestimated value rather than an actual value. The calibration curve for local heat transfer coefficients is presented using the correction factor calculated by CFD.

• 대한기계학회논문집
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• 제19권10호
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• pp.2630-2636
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• 1995

A theoretically based prediction for the determination of the subchannel friction factor at low pitch to the rod diameter ratio (P/D < 1.2) in the bare rod bundle flow has been developed. The present model assumes the validity of the Law of Wall over the entire flow area. The algebraic form of the Law of the Wall is integrated over the entire flow area and the local friction velocity variation along the rod periphery is considered in this study. The present method is applied to the rod bundles with P/D < 1.2, and the prediction results show good agreement with the available experimental data.

• 대한기계학회논문집
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• 제16권8호
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• pp.1545-1551
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• 1992

본 연구에서는 벽면의 법칙(law of the wall)을 이용하여 봉다발의 난류 마찰 계수를 예측하는 간단한 방법이 제안되었다. 이 방법은 전체 유로를 요소 유로로 나 누고 각 요소유로에 벽면의 법칙을 적용하여 마찰 계수를 구한다.이 방법을 사용 하면 마찰 계수가 간단한 산술식의 형태로 나타나므로 복잡한 형상의 봉다발의 마찰 계수도 손쉽게 구할 수 있다.

• Nuclear Engineering and Technology
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• 제28권2호
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• pp.162-174
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• 1996

서로 다른 지지격자들이 인접한 6$\times$6 핵연료 봉다발부수로내에서 국부 수력특성인자들을 레이저 유속 측정 장치인 LDV(Laser Doppler Velocimeter)를 이용하여 측정하였다. 6$\times$6 봉다발은 서로 다른 지지격자를 가진 3$\times$6 봉다발이 서로 인접하여 이룬 형상이다. 본 연구에서는 다른 형상과 다른 수력저항을 갖는 지지격자간들의 열수력적 상호작용을 규명하는데 그 목적이 있다. LDV를 이용하여 축방향 및 횡방향 속도, 난류강도 등의 측정 인자들을 측정하였다. 또한 압력강하를 측정하여 지지격자의 손실계수와 봉다발의 마찰계수를 구하였다. 수력실험결과에 근거하여 지지격자에 기인된 열혼합현상에 관한 것을 연구하였다. DNB의 정성적인 기준이라고 할 수 있는 swirl인자를 정의하고 횡방향속도 실험인자로부터 구하였다.