Modeling of molten corium-concrete interaction: Interim report

Technical Report · Tue Sep 01 00:00:00 EDT 1987

OSTI ID:6100796

Lee, M; Kazimi, M S

In the highly unlikely event of a loss of cooling accident in a LWR followed by failure of certain engineered safety features of the reactor system, the core may eventually melt due to the generation of decay heat. If the safety features of the reactor system fail to arrest the accident within the vessel, the molten core debris (corium) will fall into the reactor cavity and attack the concrete walls and floor. The heat transferred from the core melt to concrete can lead to concrete decomposition accompanied by gas generation, which along with direct heating of the atmosphere will lead to a pressure rise in the containment. The cooling rate of the core melt and the amount of gas generated by concrete decomposition will also affect the degree to which fission products may be released from the melt. A semiempirical correlation for calculating the downward heat transfer coefficient is derived based on periodic contact between the liquid pool and the underlying solid. The experimental data on interfacial heat transfer between bubble agitated immiscible layers are reviewed and a new model is proposed. The proposed downward and interfacial heat transfer models have been incorporated into an integral analysis code, CORCON/Mod1, developed by Sandia Laboratories. A simplified containment model based on thermal equilibrium among all materials within the containment is developed, and is integrated with the modified CORCON/Mod1. The combined model, called CORCELL, is used to study the impact of Corium/Concrete Interaction on containment pressurization.

• Research Organization: Massachusetts Inst. of Tech., Cambridge (USA). Dept. of Nuclear Engineering; Electric Power Research Inst., Palo Alto, CA (USA)
• OSTI ID: 6100796
• Report Number(s): EPRI-NP-5403; ON: TI88920038; TRN: 87-036602
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BWR TYPE REACTORS
MELTDOWN
CONCRETES
FLUID-STRUCTURE INTERACTIONS
CORIUM
PWR TYPE REACTORS
C CODES
CONTAINMENT
EXPERIMENTAL DATA
FISSION PRODUCTS
HEAT TRANSFER
HYDRAULICS
REACTOR SAFETY
RISK ASSESSMENT
SOURCE TERMS
ACCIDENTS
BUILDING MATERIALS
COMPUTER CODES
DATA
ENERGY TRANSFER
FLUID MECHANICS
INFORMATION
ISOTOPES
MATERIALS
MECHANICS
NUMERICAL DATA
RADIOACTIVE MATERIALS
REACTOR ACCIDENTS
REACTORS
SAFETY
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled