Spreading of molten corium in MK I geometry following vessel melt-through
Conference
·
· Trans. Am. Nucl. Soc.; (United States)
OSTI ID:5767590
For Mk I boiling water reactor severe-accident sequences in which molten corium is postulated to melt through the reactor pressure vessel (RPV) lower head, an important question concerns the relocation of the corium material that drains from the vessel. After filling the sump pits located in the pedestal concrete floor beneath the RPV, the molten corium that collects on the pedestal floor is generally free to flow through the doorway, which provides personnel access to the pedestal, and to spread out over the concrete floor in the annular region between the pedestal wall and the steel liner of the containment shell. A significant issue is whether the corium, after exiting the doorway, can spread under gravity all the way to the liner where thermal attack on the liner steel might be postulated to occur. A computer code (MELTSPREAD) has been developed to investigate the spreading dynamics and thermal interactions of a molten corium layer flowing horizontally over an ablating concrete substrate that may be initially covered with water. The principal objective is to predict, for specific conditions of corium composition, mass, and temperature, the lateral penetration of the corium that drains from a localized hole in the lower head immediately following RPV failure.
- Research Organization:
- Argonne National Lab., IL (USA)
- OSTI ID:
- 5767590
- Report Number(s):
- CONF-881011-
- Conference Information:
- Journal Name: Trans. Am. Nucl. Soc.; (United States) Journal Volume: 57
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BOILING
BUILDING MATERIALS
BWR TYPE REACTORS
CHEMICAL REACTIONS
COMPUTER CODES
CONCRETES
CONTAINERS
CONTAINMENT
CONTAINMENT SYSTEMS
CORIUM
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FILM BOILING
FINITE DIFFERENCE METHOD
FLUID MECHANICS
HEAT TRANSFER
HEATING
HYDRAULICS
ITERATIVE METHODS
LINERS
M CODES
MATERIALS
MECHANICS
MELTDOWN
MITIGATION
NUMERICAL SOLUTION
ONE-DIMENSIONAL CALCULATIONS
OXIDATION
PHASE TRANSFORMATIONS
PRESSURE VESSELS
REACTOR ACCIDENTS
REACTOR CORE DISRUPTION
REACTOR SAFETY
REACTORS
SAFETY
SUPERHEATING
WATER COOLED REACTORS
WATER MODERATED REACTORS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BOILING
BUILDING MATERIALS
BWR TYPE REACTORS
CHEMICAL REACTIONS
COMPUTER CODES
CONCRETES
CONTAINERS
CONTAINMENT
CONTAINMENT SYSTEMS
CORIUM
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FILM BOILING
FINITE DIFFERENCE METHOD
FLUID MECHANICS
HEAT TRANSFER
HEATING
HYDRAULICS
ITERATIVE METHODS
LINERS
M CODES
MATERIALS
MECHANICS
MELTDOWN
MITIGATION
NUMERICAL SOLUTION
ONE-DIMENSIONAL CALCULATIONS
OXIDATION
PHASE TRANSFORMATIONS
PRESSURE VESSELS
REACTOR ACCIDENTS
REACTOR CORE DISRUPTION
REACTOR SAFETY
REACTORS
SAFETY
SUPERHEATING
WATER COOLED REACTORS
WATER MODERATED REACTORS