Numerical predictions of natural convection in a uniformly heated pool
In the event of a core meltdown accident, one of the accident progression paths is fuel relocation to the lower reactor plenum. In the heavy water new production reactor (NPR-HWR) design the reactor cavity is flooded with water. In such a design, decay heat removal to the water in the reactor cavity and thence to the containment may be adequate to keep the reactor vessel temperature below failure limits. If this is the case, the accident progression can be arrested by retaining a coolable corium configuration in the lower reactor plenum. The strategy of reactor cavity flooding to prevent reactor vessel failure from molten corium relocation to the reactor vessel lower head has also been considered for commercial pressurized water reactors. Previously, the computer code COMMIX-LAR/P was used to determine if the heat removal rate from the molten cerium in the lower plenum to the water in the cavity was adequate to keep the reactor vessel temperature in the NPR-HWR design below failure limits. It was found that natural convection in the molten pool resulted in heat removal rates that kept the peak reactor vessel temperature about 400[degrees]C below the steel melting point. The objective of the work presented in this paper was to determine whether COMMIX adequately predicts natural convection in a pool heated by a uniform heat source. For this purpose, the experiments of free convection in a semicircular cavity of Jahn and Reeneke were analyzed with COMMIX and code predictions were compared with experimental measurements. COMMIX is a general purpose thermalhydraulics code based on finite differencing by the first order upwind scheme.
- Research Organization:
- Argonne National Lab., IL (United States)
- Sponsoring Organization:
- USDOE; USDOE, Washington, DC (United States)
- DOE Contract Number:
- W-31109-ENG-38
- OSTI ID:
- 6800169
- Report Number(s):
- ANL/RE/CP-78569; CONF-930601-17; ON: DE93012888
- Resource Relation:
- Conference: American Nuclear Society (ANS) annual meeting, San Diego, CA (United States), 20-24 Jun 1993
- Country of Publication:
- United States
- Language:
- English
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Natural convection in a uniformly heated pool
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
CORIUM
FLUID FLOW
HEAVY WATER MODERATED REACTORS
MELTDOWN
PRODUCTION REACTORS
C CODES
FINITE DIFFERENCE METHOD
HEAT TRANSFER
HYDRAULICS
NATURAL CONVECTION
REACTOR SAFETY
ACCIDENTS
CALCULATION METHODS
COMPUTER CODES
CONVECTION
ENERGY TRANSFER
FLUID MECHANICS
ITERATIVE METHODS
MASS TRANSFER
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NUMERICAL SOLUTION
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220900* - Nuclear Reactor Technology- Reactor Safety
220600 - Nuclear Reactor Technology- Research
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210400 - Power Reactors
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