Open AccessAnalysis of the thermal-hydraulic behavior resulting in early critical heat flux and evaluation of CHF correlations for the semiscale core. [PWR; LOFT]
Author(s) -
D. M. Snider
Publication year - 1977
Publication title -
osti oai (u.s. department of energy office of scientific and technical information)
Language(s) - English
Resource type - Reports
DOI - 10.2172/7114940
Subject(s) - thermal hydraulics , core (optical fiber) , pressurized water reactor , nuclear engineering , heat transfer , nuclear reactor core , environmental science , critical heat flux , thermal , nuclear reactor , core temperature , heat flux , mechanics , materials science , engineering , thermodynamics , physics , medicine , anesthesia , composite material
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