• Nuclear Engineering and Technology
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• 제37권5호
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• pp.457-468
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• 2005

An experimental study on post-CHF heat transfer has been performed with a $3\times3$ rod bundle using a vertical steam-water two-phase flow at low flow conditions. The effects of various parameters on the post-CHF heat transfer are investigated and the reasons for the parametric effects are discussed. As the heat transfer regime changes from CHF to post-CHF, the radial wall temperature distribution is changed depending on the pressure and the mass flux conditions. The superheat of the fluid increases considerably with an increase of the wall temperature (or heat flux) and with a decrease of the mass flux. This implies, indirectly, a strong thermal non-equilibrium at high wall temperature and low mass flux conditions. In order to improve the prediction accuracy of the existing post-CHF correlations, it is necessary to perform more experiments, particularly direct measurement of the vapor superheat, and to modify the correlation by considering a strong thermal non-equilibrium at low flow and low pressure conditions.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.895-910
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• 2013

The reflood test data from the rod bundle heat transfer (RBHT) test facility showed that the grids in the upper portion of the rod bundle could become wet well before the arrival of the quench front and that the sizes of liquid droplets downstream of a wet grid could not be predicted by the droplet breakup models for a dry grid. To investigate the water droplet generation from a wet grid spacer, a viscous linear temporal instability model of the water sheet issuing from the trailing edge of the grid with the surrounding steam up-flow is developed in this study. The Orr-Sommerfeld equations along with appropriate boundary conditions for the flow are solved using Chebyshev series expansions and the Tau-Galerkin projection method. The effects of several physical parameters on the water sheet oscillation are studied by determining the variation of the temporal growth rate with the wavenumber. It is found that a larger relative steam velocity to water velocity has a tendency to destabilize the water sheet with increased dynamic pressure. On the other hand, a larger ratio of steam boundary layer to the half water sheet thickness has a stabilizing effect on the water sheet oscillation. Droplet diameters downstream of the spacer grid predicted by the present model are found to compare reasonably well with the data obtained at the RBHT test facility as well as with other data recently reported in the literature.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.376-385
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• 2016

A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and $60^{\circ}C$ (equivalent to Re ~ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

• 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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.268-269
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• 2005

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1945-1954
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• 2020

Subchannel code is one of the effective simulation tools for thermal-hydraulic analysis in nuclear reactor core. In order to reduce the computational cost and improve the calculation efficiency, empirical correlation of turbulent mixing coefficient is employed to calculate the lateral mixing velocity between adjacent subchannels. However, correlations utilized currently are often fitted from data achieved in central channel of fuel assembly, which would simply neglect the wall effects. In this paper, the CFD approach based on spectral element method is employed to predict turbulent mixing phenomena through gaps in 3 × 3 bare tight lattice rod bundle and investigate the flow pulsation through gaps in different positions. Re = 5000,10000,20500 and P/D = 1.03 and 1.06 have been covered in the simulation cases. With a well verified mesh, lateral velocities at gap center between corner channel and wall channel (W-Co), wall channel and wall channel (W-W), wall channel and center channel (W-C) as well as center channel and center channel (C-C) are collected and compared with each other. The obvious turbulent mixing distributions are presented in the different channels of rod bundle. The peak frequency values at W-Co channel could have about 40%-50% reduction comparing with the C-C channel value and the turbulent mixing coefficient β could decrease around 25%. corrections for β should be performed in subchannel code at wall channel and corner channel for a reasonable prediction result. A preliminary analysis on fluctuation at channel gap has also performed. Eddy cascade should be considered carefully in detailed analysis for fluctuating in rod bundle.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.721-733
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• 2020

In nuclear engineering, the occurrence of critical heat flux (CHF) is complicated for rod bundle, and it is much more difficult to predict the CHF when it is in natural circulation under motion condition. In this paper, the dryout-type CHF is investigated for the rod bundle in a natural circulation loop under rolling motion condition based on the coupled analysis of subchannel method, a one-dimensional system analysis method and a CHF mechanism model, namely the three-fluid model for annular flow. In order to consider the rolling effect of the natural circulation loop, the subchannel model is connected to the one-dimensional system code at the inlet and outlet of the rod bundle. The subchannel analysis provides the local thermal hydraulic parameters as input for the CHF mechanism model to calculate the occurrence of CHF. The rolling motion is modeled by additional motion forces in the momentum equation. First, the calculation methods of the natural circulation and CHF are validated by a published natural circulation experiment data and a CHF empirical correlation, respectively. Then, the CHF of the rod bundle in a natural circulation loop under both the stationary and rolling motion condition is predicted and analyzed. According to the calculation results, CHF under stationary condition is smaller than that under rolling motion condition. Besides, the CHF decreases with the increase of the rolling period and angular acceleration amplitude within the range of inlet subcooling and mass flux adopted in the current research. This paper can provide useful information for the prediction of CHF in natural circulation under motion condition, which is important for the nuclear reactor design improvement and safety analysis.

• 대한기계학회논문집B
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• 제36권7호
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• pp.689-695
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• 2012

지지격자가 있는 봉다발과 축방향으로 평행한 유동에서, 봉다발 마찰계수와 지지격자 손실계수를 평가하였다. 시험부는 외경 9.5 mm, 길이 2000 mm 인 봉 25 개를 $5{\times}5$ 정사각 구조로 배열하여 제작하였으며, 봉 중심간 거리와 봉 외경의 비는 1.35 였다. 지지격자로는 plain 지지격자, split-vane 지지격자, hybrid-vane 지지격자를 이용하였다. 지지격자가 없는 봉다발의 마찰계수는 기존 상관식과 비교적 잘 일치하였다. 지지격자가 있는 봉다발 실험의 경우, hybrid-vane 지지격자에서 봉다발 마찰계수 및 지지격자 손실계수가 가장 크게 측정되었으며, 이는 지지격자의 유동단면 막음비 증가와 혼합날개 형상에 의한 유동 교란이 증가되기 때문인 것으로 판단된다. Re=$5{\times}10^5$ 조건에서 plain 지지격자, split-vane 지지격자, hybrid-vane 지지격자의 손실계수는 약 0.79, 0.80, 0.88 로 예측되었다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.169.2-169
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.194.2-194
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• 2001