Coolant Recirculation in a Pressurized Water Reactor Core under Loss-of-Coolant Accident Conditions
Author:
Affiliation:
1. Purdue University, Heat Transfer Laboratory, School of Mechanical Engineering, West Lafayette, Indiana 47907
Publisher
Informa UK Limited
Subject
Nuclear Energy and Engineering
Link
https://www.tandfonline.com/doi/pdf/10.13182/NSE88-A22323
Reference19 articles.
1. T. Y. HAN and P. NAKAYAMA, “Analysis In-Vessel Core Melt Progression, User’s Manual and Modeling Details for the PWR Heat-Up Code,” Draft Technical Report 15.1, Vol. 2, Atomic Industrial Forum (1983).
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1. Experimental examinations of the 2D-thermal-hydraulic code FRECON for single phase natural convection flows in complex structures;Nuclear Engineering and Design;1993-05
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