In-vessel flow characterization under severe accident conditions
Conference
·
OSTI ID:5991480
The purpose of this study is to provide a parametric framework for characterization of flow and heat transfer regimes and their associated phenomenological uncertainties following severe accidents using a two dimensional, heterogenous, porous media formulation. This approach extends the understanding of buoyancy-induced flow characteristics in the uncovered region of the reactor core and the upper plenum of a PWR vessel. The results of this study can be used to augment the boil-off steam flow in integrated one-dimensional severe accident codes such as the Source Team Code Package (STCP).
- Research Organization:
- Brookhaven National Lab., Upton, NY (USA); Nuclear Regulatory Commission, Washington, DC (USA)
- DOE Contract Number:
- AC02-76CH00016
- OSTI ID:
- 5991480
- Report Number(s):
- BNL-NUREG-39933; CONF-871101-37; ON: DE87011921
- Country of Publication:
- United States
- Language:
- English
Similar Records
In-vessel flow characterization under severe accident conditions
Fission product release characteristics into containment under design basis and severe accident conditions
Modelling of natural convection processes during degraded core accidents
Conference
·
Wed Dec 31 23:00:00 EST 1986
· Trans. Am. Nucl. Soc.; (United States)
·
OSTI ID:6808292
Fission product release characteristics into containment under design basis and severe accident conditions
Technical Report
·
Mon Feb 29 23:00:00 EST 1988
·
OSTI ID:6956680
Modelling of natural convection processes during degraded core accidents
Conference
·
Thu Dec 31 23:00:00 EST 1987
·
OSTI ID:5050674
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
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
COMPUTER CODES
CONTAINMENT
ENERGY TRANSFER
FLOW MODELS
FLUID MECHANICS
HEAT TRANSFER
HEATING
HYDRAULICS
LOSS OF COOLANT
MATHEMATICAL MODELS
MECHANICS
MELTDOWN
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
S CODES
SAFETY
SOURCE TERMS
STEAM
TWO-DIMENSIONAL CALCULATIONS
WATER COOLED REACTORS
WATER MODERATED REACTORS
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
COMPUTER CODES
CONTAINMENT
ENERGY TRANSFER
FLOW MODELS
FLUID MECHANICS
HEAT TRANSFER
HEATING
HYDRAULICS
LOSS OF COOLANT
MATHEMATICAL MODELS
MECHANICS
MELTDOWN
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
S CODES
SAFETY
SOURCE TERMS
STEAM
TWO-DIMENSIONAL CALCULATIONS
WATER COOLED REACTORS
WATER MODERATED REACTORS