Development of an improved mechanistic critical heat flux model for subcooled flow boiling and application to a CHF enhancement system개선된 과냉유동비등 임계열유속 모델의 개발 및 CHF 증진기구에의 적용

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An improved mechanistic model to predict a critical heat flux (CHF) over a wide operating range in subcooled or low quality forced flow boiling is proposed. The proposed CHF model is based on a new concept of the wall-attached bubble coalescence on the heated wall near the CHF. In order to derive the CHF formula in vertical round tubes with uniform heat flux, local conservation equations of mass, energy and momentum, together with appropriate constitutive relations, are solved analytically in the same manner as done by Chang and Lee (1989). The proposed model is characterized by an introduction of the frictional drag due to wall-attached bubble roughness in the momentum balance, which determines the limiting transverse interchange of mass flux crossing the interface of the wall bubbly layer and core. According to experimental results, a single layer of bubbles compactly attached on the heated wall (wall bubbly layer) is assumed to be responsible for an effective physical barrier to the heat transfer from the wall and to the liquid supply from the core near the CHF condition. It is hypothesized that CHF condition reaches at a certain void fraction in the wall bubbly layer (critical wall-void fraction) when radial thermal transport is limited by equal flows inward and outward at the interface of the wall bubbly layer and core. An empirical correlation for the critical wall-void fraction is obtained by data fitting against the KAIST CHF database. A total 5009 data points of the low quality or subcooled water flow boiling covers a wide range of operating conditions of light water reactors: 1 ≤ D ≤ 37.5 mm, 0.035 ≤ L ≤ 6 m, 450 ≤ G ≤ 7500 kg/㎡s, 2 ≤ P ≤ 20 MPa, and 0 ≤ $Δh_{sub, in}$ ≤ 1660 kJ/kg. With use of the empirical correlation, most of the experiment CHF data (about 93%) is successfully predicted within ±20% error band. The overall mean ratio of the predicted to measured CHF values is 0.99 with a standard deviation of 11.12% and a RMS error of 11.14%. The p...
Advisors
Chang, Soon-Heungresearcher장순흥researcher
Description
한국과학기술원 : 원자력공학과,
Publisher
한국과학기술원
Issue Date
1999
Identifier
156011/325007 / 000945028
Language
eng
Description

학위논문(박사) - 한국과학기술원 : 원자력공학과, 1999.8, [ xiv, 253 p. ]

Keywords

비등위기; 과냉유동비등; 임계열유속; Boiling Crisis; Subcooled Flow Boiling; Critical Heat Flux; CHF

URI
http://hdl.handle.net/10203/48899
Link
http://library.kaist.ac.kr/search/detail/view.do?bibCtrlNo=156011&flag=dissertation
Appears in Collection
NE-Theses_Ph.D.(박사논문)
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