Results and implications of the EBR-II inherent safety demonstration tests
Conference
·
OSTI ID:6770302
On April 3, 1986 two milestone tests were conducted in Experimental Breeder Reactor-2 (EBR-II). The first test was a loss of flow without scram and the second was a loss of heat sink without scram. Both tests were initiated from 100% power and in both tests the reactor was shut down by natural processes, principally thermal expansion, without automatic scram, operator intervention or the help of special in-core devices. The temperature transients during the tests were mild, as predicted, and there was no damage to the core or reactor plant structures. In a general sense, therefore, the tests plus supporting analysis demonstrated the feasibility of inherent passive shutdown for undercooling accidents in metal-fueled LMRs. The results provide a technical basis for future experiments in EBR-II to demonstrate inherent safety for overpower accidents and provide data for validation of computer codes used for design and safety analysis of inherently safe reactor plants.
- Research Organization:
- Argonne National Lab., IL (USA)
- DOE Contract Number:
- W-31109-ENG-38
- OSTI ID:
- 6770302
- Report Number(s):
- CONF-870424-10; ON: DE87006991
- Country of Publication:
- United States
- Language:
- English
Similar Records
Results and implications of the Experimental Breeder Reactor II inherent safety demonstration tests
Demonstration of passive safety features in EBR-II
Predicted and measured response of the EBR-II plant to large steam pressure changes
Journal Article
·
Wed Nov 30 23:00:00 EST 1988
· Nucl. Sci. Eng.; (United States)
·
OSTI ID:6218261
Demonstration of passive safety features in EBR-II
Conference
·
Wed Dec 31 23:00:00 EST 1986
·
OSTI ID:6174616
Predicted and measured response of the EBR-II plant to large steam pressure changes
Conference
·
Wed Dec 31 23:00:00 EST 1986
·
OSTI ID:6782044
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
AFTER-HEAT REMOVAL
BREEDER REACTORS
DATA
EBR-2 REACTOR
ENERGY TRANSFER
EPITHERMAL REACTORS
EXPERIMENTAL DATA
EXPERIMENTAL REACTORS
FAILURE MODE ANALYSIS
FAST REACTORS
FBR TYPE REACTORS
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INFORMATION
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
LOSS OF FLOW
MECHANICS
NUMERICAL DATA
OPERATION
POWER REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORES
REACTOR OPERATION
REACTOR SAFETY EXPERIMENTS
REACTOR SHUTDOWN
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
SCRAM
SHUTDOWNS
SODIUM COOLED REACTORS
SYSTEM FAILURE ANALYSIS
SYSTEMS ANALYSIS
THERMAL ANALYSIS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
AFTER-HEAT REMOVAL
BREEDER REACTORS
DATA
EBR-2 REACTOR
ENERGY TRANSFER
EPITHERMAL REACTORS
EXPERIMENTAL DATA
EXPERIMENTAL REACTORS
FAILURE MODE ANALYSIS
FAST REACTORS
FBR TYPE REACTORS
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INFORMATION
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
LOSS OF FLOW
MECHANICS
NUMERICAL DATA
OPERATION
POWER REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORES
REACTOR OPERATION
REACTOR SAFETY EXPERIMENTS
REACTOR SHUTDOWN
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
SCRAM
SHUTDOWNS
SODIUM COOLED REACTORS
SYSTEM FAILURE ANALYSIS
SYSTEMS ANALYSIS
THERMAL ANALYSIS