The MELTSPREAD-1 computer code for the analysis of transient spreading in containments
Conference
·
OSTI ID:6604685
A one-dimensional, multicell, Eulerian finite difference computer code (MELTSPREAD-1) has been developed to provide an improved prediction of the gravity driven spreading and thermal interactions of molten corium flowing over a concrete or steel surface. In this paper, the modeling incorporated into the code is described and the spreading models are benchmarked against a simple dam break'' problem as well as water simulant spreading data obtained in a scaled apparatus of the Mk I containment. Results are also presented for a scoping calculation of the spreading behavior and shell thermal response in the full scale Mk I system following vessel meltthrough. 24 refs., 15 figs.
- Research Organization:
- Argonne National Lab., IL (USA)
- Sponsoring Organization:
- EPRI
- DOE Contract Number:
- W-31109-ENG-38
- OSTI ID:
- 6604685
- Report Number(s):
- CONF-901101-30; ON: DE90017695; CNN: RP 3047-2
- Country of Publication:
- United States
- Language:
- English
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The MELTSPREAD-1 computer code for the analysis of transient spreading in containments
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Conference
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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
BWR TYPE REACTORS
C CODES
COMPUTER CALCULATIONS
COMPUTER CODES
CONTAINMENT
CORIUM
ENERGY TRANSFER
FLOW MODELS
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
M CODES
MATHEMATICAL MODELS
MECHANICS
MELTDOWN
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY
REACTORS
SAFETY
THERMAL ANALYSIS
TRANSIENTS
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
BWR TYPE REACTORS
C CODES
COMPUTER CALCULATIONS
COMPUTER CODES
CONTAINMENT
CORIUM
ENERGY TRANSFER
FLOW MODELS
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
M CODES
MATHEMATICAL MODELS
MECHANICS
MELTDOWN
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY
REACTORS
SAFETY
THERMAL ANALYSIS
TRANSIENTS
WATER COOLED REACTORS
WATER MODERATED REACTORS