• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.480-486
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• 1995

Turbulent flow field in a subchannel of bare rod bundles has been numerically simulated using the control volume based finite element method. Launder & Ying model of Reynolds stress and Lam & Bremhorst low-Reynolds number model are implemented in k-$\varepsilon$ equations and momentum equations. Secondary flows are simulated using the stream function and vorticity approach. The control volume based finite element method enable to use the upwind scheme (donor cell scheme). Sensitivity of the constants in the models are studied, and proper values are found to get the close result to the measured flow distributions.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.75-81
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• 1996

COBRA-TF is a two fluid, three field subchannel code. Three fields are continuous vapor, continuous liquid and droplet. Some assessments are conducted to validate the related models and to estimate a code ability through dryout and post-CHF experiment in a tube and DNB test in rod bundles. It turned out form dryout and post-CHF experiment that the predicted dryout locations and wall temperature profiles are in close agreement with the experiments. On the other hand, DNB prediction of COBRA-TF are performed for two kinds of rod bundles along with EPRI CHF correlation. To estimate its performance COBRA-IV of homogeneous model is also run for the same data. The results say that COBRA-TF/EPRI is better in DNB prediction than COBRA-IV/EPRI. In addition the thermal-hydraulic behaviors due to the different two-phase flow models are presented at the condition of CHF.

• Nuclear Engineering and Technology
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• 제31권1호
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• pp.104-115
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• 1999

The turbulent mixing rate is a very important variable in the thermal-hydraulic design of nuclear reactors. In this study, the turbulent mixing rate the fluid flows through rod bundles is estimated with the scale analysis on the flow pulsation phenomenon. Based upon the assumption that the turbulent mixing is composed of molecular motion, isotropic turbulent motion (turbulent motion without the flow pulsation), and How pulsation, the scale relation for the mixing is derived as a function of P/D, Re, and Pr. The derived scale relation is compared with published experimental results and shows good agreements. Since the scale relation is applicable to various Prandtl number fluid flows, it is expected to be useful for the thermal-hydraulic analysis of liquid metal coolant reactors as well as of moderate Prandtl number coolant reactors.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.284-292
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• 2015

A numerical calculation with the commercial computational fluid dynamics code CFX-14.0 was conducted for a test facility simulating the Canadian deuterium uranium moderator thermal-hydraulic. Two kinds of moderator thermal-hydraulic tests at Stern Laboratories Inc. were performed in the full geometric configuration of the Canadian deuterium uranium moderator circulating vessel, which is called a calandria tank, housing a matrix of horizontal rod bundles simulating calandria tubes. The first of these tests is the pressure drop measurement of a cross flow in the horizontal rod bundles. The other is the local temperature measurement on the cross section of the horizontal cylinder vessel simulating the calandria system. In the present study, the full geometric details of the calandria tank are incorporated in the grid generation of the computational domain to which the boundary conditions for each experiment are applied. The numerical solutions are reviewed and compared with the available test data.

• 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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 추계학술대회논문집B
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• pp.3-8
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• 2001

For the successful design of nuclear reactor, it is very important to investigate thermal-hydraulic characteristics of fuel rod bundle. Fluid flow and heat transfer in the non-circular cross-section of nuclear fuel rod bundle are different from those found in common circular tube. And complex three dimensional flow including secondary and vortex flow, is formed around the bundles. The purpose of this research is to examine how geometries and flow conditions affect heat transfer in fuel rod bundle. Design data for nuclear fuel rod bundle and structure are surveyed, and $3{\times}3$ sub-channel model is adopted in this study. Computational results are compared with the heat transfer data measured by naphthalene sublimation method, and numerical analysis and evaluation are performed at various design conditions and flow conditions.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.156-156
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• 1999

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.36-48
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• 2022

As a valid numerical method to obtain a high-resolution result of a flow field, computational fluid dynamics (CFD) have been widely used to study coolant flow and heat transfer characteristics in fuel rod bundles. However, the time-consuming, iterative calculation of Navier-Stokes equations makes CFD unsuitable for the scenarios that require efficient simulation such as sensitivity analysis and uncertainty quantification. To solve this problem, a reduced-order model (ROM) based on proper orthogonal decomposition (POD) and machine learning (ML) is proposed to simulate the flow field efficiently. Firstly, a validated CFD model to output the flow field data set of the rod bundle is established. Secondly, based on the POD method, the modes and corresponding coefficients of the flow field were extracted. Then, an deep feed-forward neural network, due to its efficiency in approximating arbitrary functions and its ability to handle high-dimensional and strong nonlinear problems, is selected to build a model that maps the non-linear relationship between the mode coefficients and the boundary conditions. A trained surrogate model for modes coefficients prediction is obtained after a certain number of training iterations. Finally, the flow field is reconstructed by combining the product of the POD basis and coefficients. Based on the test dataset, an evaluation of the ROM is carried out. The evaluation results show that the proposed POD-ROM accurately describe the flow status of the fluid field in rod bundles with high resolution in only a few milliseconds.

• 한국조명전기설비학회:학술대회논문집
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• 한국조명전기설비학회 2000년도 학술대회논문집
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• pp.179-182
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• 2000

This paper investigates characteristics of spiral charge for charge simulation method to calculate electric fields of spiral conductor bundles with spiral rods in 765 kV transmission line. We finds the simulating spiral charge as constant charge density give less potential calculation error than sinusoidal charge density. When a spiral rod is simulated as spiral charge, we also knows two spiral charge can simulate spiral rod best.

• 조명전기설비학회논문지
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• 제15권4호
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• pp.31-36
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• 2001

본 논문은 765[kV] 송전선로의 풍소음 저감용 spiral rod의 전계를 계산하기 위해서 3차원 전하중첩법용 나선 전하의 특성에 관해 연구하였다. 본 연구에서 나선전하를 정현 함수 전하로 모의하는 것 보다 상수전하로 모의한 것이 전위오차가 적었고, 상수전하를 한 개로 모의하는 것 보다 2개를 배치하여 계산했을 때 spiral rod의 전하를 가장 잘 모의할 수 있는 것으로 나타났다.