Results and implications of the EBR-II inherent safety demonstration tests
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
- Resource Relation:
- Conference: American Nuclear Society international meeting on advances in reactor physics, mathematics and computation, Paris, France, 27 Apr 1987
- Country of Publication:
- United States
- Language:
- English
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
EBR-2 REACTOR
REACTOR OPERATION
REACTOR SAFETY EXPERIMENTS
AFTER-HEAT REMOVAL
EXPERIMENTAL DATA
FAILURE MODE ANALYSIS
HEAT TRANSFER
HYDRAULICS
LOSS OF FLOW
REACTOR CORES
SCRAM
THERMAL ANALYSIS
ACCIDENTS
BREEDER REACTORS
DATA
ENERGY TRANSFER
EPITHERMAL REACTORS
EXPERIMENTAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FLUID MECHANICS
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LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
MECHANICS
NUMERICAL DATA
OPERATION
POWER REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SHUTDOWN
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
SHUTDOWNS
SODIUM COOLED REACTORS
SYSTEM FAILURE ANALYSIS
SYSTEMS ANALYSIS
220900* - Nuclear Reactor Technology- Reactor Safety
220600 - Nuclear Reactor Technology- Research
Test & Experimental Reactors