Abstract
In this study, a critical heat flux (CHF) experiment using R-134a as working fluid was carried out in a vertical single rod geometry. A unique feature of this experiment are high speed visualization studies obtained at a viewing port located at the CHF location. The 9.5 mm O.D. heated rod was placed in a 19 x19mm square channel to simulate the Pressurized Water Reactors (PWRs) fuel lattice. Test conditions were selected and scaled to cover a wide range of water conditions relevant to PWRs. The CHF database, experimental trends and comparisons to existing CHF prediction methods are presented. The visualization results of the flow boiling at the CHF location and the CHF mechanism are discussed with a particular emphasis on the impact of grid spacers on the boiling phenomena. Quantitative analysis of bubble layer thickness is carried out, including the relationship between bubble layer thickness and CHF.
Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have
Schedule a callDisclaimer: All third-party content on this website/platform is and will remain the property of their respective owners and is provided on "as is" basis without any warranties, express or implied. Use of third-party content does not indicate any affiliation, sponsorship with or endorsement by them. Any references to third-party content is to identify the corresponding services and shall be considered fair use under The CopyrightLaw.
