SCDAP severe core-damage studies: BWR ATWS and PWR station blackout
Conference
·
OSTI ID:5624615
The Severe Accident Sequence Analysis (SASA) Program, sponsored by the US Nuclear Regulatory Commission (NRC), is addressing a number of accident scenarios that potentially pose a health hazard to the public. Two of the scenarios being analyzed in detail at the Idaho National Engineering Laboratory (INEL) are the station blackout at the Bellefonte nuclear plant and the anticipated transient without scram (ATWS) at the Browns Ferry-1 plant. The INEL analyses of the station blackout and ATWS have been divided into four parts, which represent the sequence being followed in this study. First, the evaluation of long term irradiation effects prior to the station blackout or ATWS was conducted using the FRAPCON-2 fuel rod behavior code; second, the reactor primary and secondary coolant system behavior is being analyzed with the RELAP5 code; third, the degradation of the core is being analyzed with the SCDAP code; and finally, the containment building response is being analyzed with the CONTEMPT code. This paper addresses only the SCDAP/MODO degraded core analyses for both the station blackout and ATWS scenarios.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 5624615
- Report Number(s):
- EGG-M-21483; CONF-8310143-33; ON: DE84002113
- Country of Publication:
- United States
- Language:
- English
Similar Records
Pressurized-water-reactor station blackout
Thermal-hydraulic and core damage analysis of the station blackout transient in pressurized water reactors
Overview of BWR Severe Accident Sequence Analyses at Oak Ridge National Laboratory
Conference
·
Fri Dec 31 23:00:00 EST 1982
·
OSTI ID:5624652
Thermal-hydraulic and core damage analysis of the station blackout transient in pressurized water reactors
Conference
·
Sat Dec 31 23:00:00 EST 1983
·
OSTI ID:6315107
Overview of BWR Severe Accident Sequence Analyses at Oak Ridge National Laboratory
Conference
·
Fri Dec 31 23:00:00 EST 1982
·
OSTI ID:5691959
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
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
ACCIDENTS
ATWS
BWR TYPE REACTORS
CHEMICAL REACTIONS
DAMAGE
ELECTRONIC EQUIPMENT
ELEMENTS
ENERGY TRANSFER
EQUIPMENT
FAILURES
FLUID MECHANICS
FUEL CANS
HEAT TRANSFER
HYDRAULICS
HYDROGEN
MECHANICS
NONMETALS
OXIDATION
POWER SUPPLIES
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORES
REACTOR SAFETY
REACTORS
SAFETY
STRESSES
TEMPERATURE GRADIENTS
THERMAL STRESSES
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
ACCIDENTS
ATWS
BWR TYPE REACTORS
CHEMICAL REACTIONS
DAMAGE
ELECTRONIC EQUIPMENT
ELEMENTS
ENERGY TRANSFER
EQUIPMENT
FAILURES
FLUID MECHANICS
FUEL CANS
HEAT TRANSFER
HYDRAULICS
HYDROGEN
MECHANICS
NONMETALS
OXIDATION
POWER SUPPLIES
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORES
REACTOR SAFETY
REACTORS
SAFETY
STRESSES
TEMPERATURE GRADIENTS
THERMAL STRESSES
WATER COOLED REACTORS
WATER MODERATED REACTORS