The reflooding phase after a loss-of-coolant accident in an advanced pressurized water reactor
Journal Article
·
· Nucl. Technol.; (United States)
OSTI ID:5913488
Calculations of the reflooding phase during a loss-of-coolant accident (LOCA) have been performed for two homogeneous advanced pressurized water reactors (APWRs) with a wide (pitch-to-diameter (p/d) ratio = 1.2) and a tighter (p/d = 1.123) fuel rod lattice as well as for a reference 1300-MW (electric) pressurized water reactor (PWR). The FLUT computer code, developed by the Gessellschaft fur Reaktorsicherheit in Garching for the reflooding phase of a PWR, has been improved. A new criterion for the determination of the onset of the upper quench front and a new water droplet model for the dispersed flow film boiling have been introduced in the code, as well as new friction factor correlations more suitable for the core geometry of an APWR. Finally, the interfacial drag coefficients between steam and water are not independent of the geometry as in FLUT, but rather the flow channel geometry is taken into account.
- Research Organization:
- 3587000
- OSTI ID:
- 5913488
- Journal Information:
- Nucl. Technol.; (United States), Journal Name: Nucl. Technol.; (United States) Vol. 84:1; ISSN NUTYB
- Country of Publication:
- United States
- Language:
- English
Similar Records
Loss-of-coolant-accident and anticipated transient without scram calculations for homogeneous and heterogeneous advanced pressurized water reactors
Experimental investigations on the reflooding and deformation behavior of an advanced pressurized water reactor tight-lattice fuel rod bundle in a loss-of-coolant accident
Dispersed-flow heat transfer during reflood in a pressurized water reactor after a large-break loss-of-coolant accident
Journal Article
·
Thu Dec 31 23:00:00 EST 1987
· Nucl. Technol.; (United States)
·
OSTI ID:7010741
Experimental investigations on the reflooding and deformation behavior of an advanced pressurized water reactor tight-lattice fuel rod bundle in a loss-of-coolant accident
Journal Article
·
Thu Dec 31 23:00:00 EST 1987
· Nucl. Technol.; (United States)
·
OSTI ID:7226391
Dispersed-flow heat transfer during reflood in a pressurized water reactor after a large-break loss-of-coolant accident
Conference
·
Tue Dec 31 23:00:00 EST 1985
· Trans. Am. Nucl. Soc.; (United States)
·
OSTI ID:6907166
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220400 -- Nuclear Reactor Technology-- Control Systems
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COMPUTER CODES
DRAG
DROPLET MODEL
F CODES
FUEL ELEMENTS
FUEL RODS
FUEL-CLADDING INTERACTIONS
GEOMETRY
LOSS OF COOLANT
MATHEMATICAL MODELS
MATHEMATICS
NUCLEAR MODELS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORES
REACTOR SAFETY EXPERIMENTS
REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220400 -- Nuclear Reactor Technology-- Control Systems
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COMPUTER CODES
DRAG
DROPLET MODEL
F CODES
FUEL ELEMENTS
FUEL RODS
FUEL-CLADDING INTERACTIONS
GEOMETRY
LOSS OF COOLANT
MATHEMATICAL MODELS
MATHEMATICS
NUCLEAR MODELS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORES
REACTOR SAFETY EXPERIMENTS
REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS