• Title/Summary/Keyword: 열분배모델

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Numerical Study of Low-pressure Subcooled Flow Boiling in Vertical Channels Using the Heat Partitioning Model (열분배모델을 이용한 수직유로에서의 저압 미포화비등 해석)

  • Lee, Ba-Ro;Lee, Yeon-Gun
    • Transactions of the Korean Society of Mechanical Engineers B
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    • v.40 no.7
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    • pp.457-470
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    • 2016
  • Most CFD codes, that mainly adopt the heat partitioning model as the wall boiling model, have shown low accuracies in predicting the two-phase flow parameters of subcooled boiling phenomena under low pressure conditions. In this study, a number of subcooled boiling experiments in vertical channels were analyzed using a thermal-hydraulic component code, CUPID. The prediction of the void fraction distribution using the CUPID code agreed well with experimental data at high-pressure conditions; whereas at low-pressure conditions, the predicted void fraction deviated considerably from measured ones. Sensitivity tests were performed on the submodels for major parameters in the heat partitioning model to find the optimized sets of empirical correlations suitable for low-pressure subcooled flow boiling. The effect of the K-factor on the void fraction distribution was also evaluated.

Visualization Experiment for Nucleate Boiling Bubble Motion on a Horizontal Tube Heater Fabricated with Flexible Circuit Board (연성회로기판 기반 수평전열관 표면의 비등기포거동 가시화 실험 연구)

  • Kim, Jae Soon;Kim, Yu-Na;Park, Goon-Cherl;Cho, Hyoung Kyu
    • Journal of the Korean Society of Visualization
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    • v.14 no.2
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    • pp.52-60
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    • 2016
  • The Passive Auxiliary Feedwater System(PAFS) is one of the advanced safety concepts adopted in the Advanced Power Reactor Plus(APR+). To validate the operational performance of the PAFS, detailed understanding of a boiling heat transfer on horizontal tube outside is of great importance. Especially, in the mechanistic boiling heat transfer model, it is important to visualize the phenomena but there are some limitations with conventional experimental approaches. In the present study, we devised a heater based on the Flexible Printed Circuit Board (FPCB) for a more comprehensive visualization and subsequently, a digital image processing technique for the bubble motion measurement was established. Using the measurement technique, important parameters of the nucleate boiling are analyzed.

Preliminary Study for the Development of Optimum Fuel Contact Conductance Model (최적 핵연료 접촉 열전도도 모델 개발을 위한 예비 연구)

  • Yang, Yong-Sik;Shin, Chang-Hwan
    • Proceedings of the KSME Conference
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    • 2007.05b
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    • pp.2488-2493
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    • 2007
  • A gap conductance is very important factor which can affect nuclear fuel temperature. Especially, in case of an annular fuel, a gap conductance effect can lead an unexpected heat split phenomena which is caused by a large difference of an inner and outer gap conductance. The gap conductance mechanism is very complicated behavior due to the its strong dependency on microscopic factors such as a contact surface roughness, local contact pressure and local temperature. In this paper, for the decision of test temperature and pressure range, a procedure and calculation results of in-reactor fuel temperature and pressure analysis are summarized which can be applied to test equipment design and determination of test matrix. Based upon analysis results, it is concluded that the minimum and maximum test temperature are $300^{\circ}C$ and $530^{\circ}C$ respectively, and the maximum pellet/cladding interfacial contact pressure should be observed up to 45MPa.

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