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dc.contributor.authorChen, Yuzhou
dc.contributor.authorZhao, Minfu
dc.contributor.authorBi, Keming
dc.contributor.authorYang, Bin
dc.contributor.authorZhang, Dongxu
dc.contributor.authorDu, Kaiwen
dc.date.accessioned2019-02-28T11:05:21Z
dc.date.available2019-02-28T11:05:21Z
dc.date.copyright12/4/2017 12:00:00 AM
dc.date.issued2018
dc.identifier.issn2332-8983
dc.identifier.otherners_004_01_011006.pdf
dc.identifier.urihttp://yetl.yabesh.ir/yetl1/handle/yetl/4252549
dc.description.abstractCritical heat flux (CHF) experiment with uniform heating was performed in a tube of 8.2 mm in inner diameter and 2.4 m in heated length. The water flowed upward through the test section. The pressure covered the range from 8.6 to 20.8 MPa, mass flux 1157 to 3776 kg/m2s, inlet quality −2.79 to −0.08 (subcooling 19–337 °C), and local quality −0.97 to 0.53. For the pressure close to the near-critical point, the CHF decreased substantially with the pressure increasing. For the subcooling larger than a certain value, the CHF was related to the local condition. But for low subcooling and saturated condition, the CHF was related to the total power. The present results were in agreement with the previous experiment for the same local subcooled condition. Based on the present experimental results with subcooled and saturated conditions an empirical relation of the CHF was presented.
dc.publisherThe American Society of Mechanical Engineers (ASME)
dc.titleCritical Heat Flux of Flowing Water in Tube for Pressure Up to Near Critical Point—Experiment and Prediction
dc.typeJournal Paper
dc.journal.volume4
dc.journal.issue1
dc.journal.titleJournal of Nuclear Engineering and Radiation Science
dc.identifier.doi10.1115/1.4038215
dc.journal.fristpage11006
dc.journal.lastpage011006-5
dc.treeJournal of Nuclear Engineering and Radiation Science:;2018:;volume( 004 ):;issue: 001
dc.contenttypeFulltext


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