Direct containment heating and aerosol generation during high-pressure-melt expulsion experiments
Conference
·
· Trans. Am. Nucl. Soc.; (United States)
OSTI ID:5862125
Severe nuclear plant accidents can involve the degradation of the reactor core while the primary coolant system remains pressurized. Molten fuel reaching the lower head of the reactor pressure vessel (RPV) may attack and fail the instrument guide tube penetrations, allowing the tube to be expelled from the vessel. The resulting aperture allows the molten fuel to be ejected into the cavity, followed by the blowdown of the contents of the primary system (high-pressure-melt ejection). Entrainment of the core debris in the cavity by the blowdown gases may cause high-temperature fuel particles to be carried into the containment building. Energy exchange between the particles and the atmosphere may cause heating and pressurizing of the containment (direct containment heating (DCH)). The complex phenomena associated with direct containment heating accident sequences are not well understood. This work describes a series of four experiments that have been performed to study and quantify the processes involved. The data from the experiments are used to guide the development of computer models to describe the response of containments under accident conditions.
- OSTI ID:
- 5862125
- 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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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
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220900* -- Nuclear Reactor Technology-- Reactor Safety
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CORIUM
DISPERSIONS
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K CODES
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NATIONAL ORGANIZATIONS
OXIDATION
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REACTOR CORES
REACTOR INSTRUMENTATION
REACTOR SAFETY
REACTOR SAFETY EXPERIMENTS
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SCALE MODELS
SIZE
SOLS
STRUCTURAL MODELS
TEST FACILITIES
TUBES
US AEC
US DOE
US ERDA
US ORGANIZATIONS
VOLATILITY
WATER COOLED REACTORS
WATER MODERATED REACTORS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
42 ENGINEERING
420400 -- Engineering-- Heat Transfer & Fluid Flow
ACCIDENTS
AEROSOLS
BLOWDOWN
BWR TYPE REACTORS
C CODES
CHEMICAL REACTIONS
COLLOIDS
COMPUTER CODES
CONTAINERS
CONTAINMENT
CONTAINMENT SYSTEMS
COOLING SYSTEMS
CORIUM
DISPERSIONS
EFFICIENCY
ENERGY SYSTEMS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID MECHANICS
HEAT TRANSFER
HEATING
HYDRAULICS
K CODES
MECHANICS
NATIONAL ORGANIZATIONS
OXIDATION
PARTICLE SIZE
PRESSURE VESSELS
PRESSURIZING
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
RADIOACTIVE AEROSOLS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR CORE DISRUPTION
REACTOR CORES
REACTOR INSTRUMENTATION
REACTOR SAFETY
REACTOR SAFETY EXPERIMENTS
REACTORS
SAFETY
SANDIA LABORATORIES
SCALE MODELS
SIZE
SOLS
STRUCTURAL MODELS
TEST FACILITIES
TUBES
US AEC
US DOE
US ERDA
US ORGANIZATIONS
VOLATILITY
WATER COOLED REACTORS
WATER MODERATED REACTORS