Experimental study for critical heat flux in 2x2 rod bundles at high pressure conditions

Critical heat flux (CHF) phenomena has been a topic of study for over fifty years yet remains a serious concern across industries and especially in water-cooled nuclear reactor design. CHF data in the high-pressure range is limited, and the effects of bundle geometry and non-uniform heat flux profil...

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Veröffentlicht in:	Nuclear engineering and design 2020-06, Vol.365 (C)
Hauptverfasser:	Lyons, Kathleen, Lee, Donghwi, Anderson, Mark
Format:	Artikel
Sprache:	eng
Schlagworte:	CHF correlation Heat transfer NUCLEAR PHYSICS AND RADIATION PHYSICS Nuclear Science & Technology Pressurized water reactor Rod bundle Water-cooled nuclear reactor
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