One of the phenomena involved in a loss-of-coolant accident in a pressurized water reactor may be lower plenum voiding. This might occur during the blowdown phase after a cold-leg break in the primary coolant circuit. Steam generated in the reactor core may flow out of the bottom of the reactor core, turn in the lower plenum of the vessel, in a direction countercurrent to the emergency core coolant flow, and escape via the break. If its velocity is high enough, this steam may sweep water from the bottom (lower plenum) of the reactor vessel. Emergency coolant added to the vessel may also be carried out by the escaping steam and thus the reflooding of the core would be delayed. This paper describes a study of two-phase hydrodynamics associated with lower plenum voiding. Several geometrical configurations were tested at three different scales, using air to simulate the steam. Comparisons were made with data obtained by other researchers.
Skip Nav Destination
Article navigation
Research Papers
Lower Plenum Voiding Available to Purchase
D. Bharathan,
D. Bharathan
Thayer School of Engineering, Darmouth College, Hanover, N.H. 03755
Search for other works by this author on:
G. B. Wallis,
G. B. Wallis
Thayer School of Engineering, Darmouth College, Hanover, N.H. 03755
Search for other works by this author on:
H. J. Richter
H. J. Richter
Thayer School of Engineering, Darmouth College, Hanover, N.H. 03755
Search for other works by this author on:
D. Bharathan
Thayer School of Engineering, Darmouth College, Hanover, N.H. 03755
G. B. Wallis
Thayer School of Engineering, Darmouth College, Hanover, N.H. 03755
H. J. Richter
Thayer School of Engineering, Darmouth College, Hanover, N.H. 03755
J. Heat Transfer. Aug 1982, 104(3): 479-486 (8 pages)
Published Online: August 1, 1982
Article history
Received:
July 30, 1981
Online:
October 20, 2009
Citation
Bharathan, D., Wallis, G. B., and Richter, H. J. (August 1, 1982). "Lower Plenum Voiding." ASME. J. Heat Transfer. August 1982; 104(3): 479–486. https://doi.org/10.1115/1.3245118
Download citation file:
Get Email Alerts
Cited By
Related Articles
External Hazard Coinciding With Small Break LOCA—Thermohydraulic Calculation With System Code ATHLET
ASME J of Nuclear Rad Sci (April,2020)
Methodology for Calculating Minor Radioactive Releases From VVER 1000 Using TRACE Code
ASME J of Nuclear Rad Sci (April,2021)
PBMR-A Future Failsafe Gas Turbine Nuclear Power Plant?
Mechanical Engineering (August,2011)
An Analytical Model for Prediction of Two-Phase (Noncondensable) Flow Pump Performance
J. Fluids Eng (March,1985)
Related Proceedings Papers
Related Chapters
Modeling of SAMG Operator Actions in Level 2 PSA (PSAM-0164)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Insights and Results of the Shutdown PSA for a German SWR 69 Type Reactor (PSAM-0028)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Link between Level 2 PSA and Off-Site Emergency Preparedness (PSAM-0363)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)