Fluid-to-Fluid modelling of CHF at High-Pressure subcooled water conditions-Reference-Cited by-同舟云学术

Fluid-to-Fluid modelling of CHF at High-Pressure subcooled water conditions

Published:2022-01 Issue: Volume:386 Page:111577
ISSN:0029-5493
Container-title:Nuclear Engineering and Design
language:en
Short-container-title:Nuclear Engineering and Design

Author:

Guo Junliang,Kong Huanjun,Gui Miao,Liu Yang,Shan Jianqiang,Leung Laurence K.H.

Funder

National Key Research and Development Program of China

Ministry of Science and Technology of the People's Republic of China

Publisher

Elsevier BV

Subject

Mechanical Engineering,Waste Management and Disposal,Safety, Risk, Reliability and Quality,General Materials Science,Nuclear Energy and Engineering,Nuclear and High Energy Physics

Reference32 articles.

1. An improved version of the generalized correlation of critical heat flux for the forced convective boiling in uniformly heated vertical tubes;Katto;Int. J. Heat Mass Transf.,1984

2. Bowring, R.W. A simple but accurate round tube uniform heat flux, dry-out correlation over the pressure range, 0.7-12MN/m2 (100-2500psia), AEEW-R789, (1972).

3. Critical heat flux (CHF) for water flow in tubes—II.: subcooled CHF correlations;Hall;Int. J. Heat Mass Transf.,2000

4. A generalized graphical method for predicting CHF in uniformly heated vertical tubes;Shah;Int. J. Heat Mass Transf.,1979

5. A mechanistic critical heat flux model for subcooled flow boiling based on local bulk flow conditions;Lee;Int. J. Multiph. Flow,1988

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