2003 Fiscal Year Final Research Report Summary
A Study on turbulence structure and critical heat flux in boiling two-phase flow
| Project/Area Number |
14350108
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| Research Category |
Grant-in-Aid for Scientific Research (B)
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| Allocation Type | Single-year Grants |
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| Section | 一般 |
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| Research Field |
Thermal engineering
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| Research Institution | Osaka University |
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Principal Investigator |
KATAOKA Isao Osaka University, Dept. of Mechanophysics Eng., Professor, 大学院・工学研究科, 教授 (80093219)
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| Co-Investigator(Kenkyū-buntansha) |
YOSHIDA Kenji Osaka Univ., Dept. of Mechanophysics Eng., Research Assistant, 大学院・工学研究科, 助手 (50314365)
MATSUMOTO Tadayoshi Osaka Univ., Dept. of Mechanophysics Eng., Research Assistant, 大学院・工学研究科, 助手 (10294018)
OKAWA Tomio Osaka Univ., Dept. of Associate Mechanophysics Eng., Professor, 大学院・工学研究科, 助教授 (20314362)
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| Project Period (FY) |
2002 – 2003
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| Keywords | Critical Heat Flux / Turbulence Structure / Gas Liquid Two-Phase Flow / Forced Convective Boiling / Bubble Dispersion Coefficient / Void Fraction / Multi-Dimensional Flow / Interfacial Structure |
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| Research Abstract |
In order to establish general predictive method of CHF based on turbulence mechanisms of boiling two-phase flow, investigation was carried out on detailed behavior of boiling bubbles at heated wall and turbulence field, mechanisms of void fraction distribution in bubbly flow in various flow geometries and film behavior and turbulence structure in gas core in annular two-phase flow in 2002. In 2003, more detailed modeling and analyses were carried out on turbulence structure in boiling two-phase flow and prediction of CHF, including actual operating conditions of nuclear reactor.
In actual operating conditions of nuclear reactors (high temperature and high pressure), analyses of turbulence structure and prediction of CHF were carried out for fuel assembly of nuclear reactor. Experimental results showed that peak void fraction was observed near wall for lower pressure (5Mpa) while peak void fraction appears in the center of subchannel for higher pressure (15Mpa). In order to explain these phenomena, mechanistic model of bubble dispersion coefficient was developed considering the effect of pressure and modified correlation of bubble dispersion coefficient was derived. Using this correlation, void fraction distribution and CHF for rod bundle were predicted.. The prediction agreed well with the experimental data of void fraction distribution and CHF for single channel and 5x5rod bundle.
In consideration of nonuniformity of heat flux in actual conditions of nuclear reactor, dryout heat flux was predicted in annular two-phase flow. In the prediction, the correlations of entrainment rate, deposition rate, quality and droplet fraction at the onset of annular flow and liquid film thickness at dryout were modeled and evaluated in details. Using these correlations, dryout heat fluxes ware accurately predicted for wide ranges of pressure, equivalent diameter and length of flow channel and liquid flow rate under the conditions of uniform and nonuniform heat flux distributions.
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