Effect of primary pump coastdown on unprotected loss-of-flow transients in EBR-II
Conference
·
OSTI ID:5986422
A loss-of-flow (LOF) accident of considerable interest in current LMFBR safety study is a total loss of pumping power coupled with a failure of the reactor shutdown system (RSS). Confirming predictions of this type of unprotected transient is the primary purpose of the Shutdown Heat Removal Tests (SHRT) presently scheduled for June 1985 in the Experimental Breeder Reactor II (EBR-II). The tests are also intended to validate predictions of the maximum operating power at which the reactor can safely sustain a total loss of power (station blackout) and a failure of RSS. An extensive series of tests was successfully completed in June of 1984 which investigated the reactor shutdown cooling capability following protected LOF events and the inherent reactivity feedback characteristics in EBR-II in preparation for making predictions of the upcoming unprotected tests. This paper discusses details of the unprotected LOF transient analysis, the validation of the EBR-II reactivity feedback modeling, and the significance of primary pump coastdown characteristics on peak reactor temperature.
- Research Organization:
- Argonne National Lab., IL (USA)
- DOE Contract Number:
- W-31109-ENG-38
- OSTI ID:
- 5986422
- Report Number(s):
- CONF-850420-2; ON: DE85002148
- Country of Publication:
- United States
- Language:
- English
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· Trans. Am. Nucl. Soc.; (United States)
·
OSTI ID:5852258
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210500 -- Power Reactors
Breeding
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BREEDER REACTORS
COOLING SYSTEMS
EBR-2 REACTOR
ENERGY SYSTEMS
EPITHERMAL REACTORS
EXPERIMENTAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FEEDBACK
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
LOSS OF FLOW
POWER REACTORS
PRIMARY COOLANT CIRCUITS
PUMPS
REACTIVITY
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
RESEARCH AND TEST REACTORS
SODIUM COOLED REACTORS
210500 -- Power Reactors
Breeding
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BREEDER REACTORS
COOLING SYSTEMS
EBR-2 REACTOR
ENERGY SYSTEMS
EPITHERMAL REACTORS
EXPERIMENTAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FEEDBACK
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
LOSS OF FLOW
POWER REACTORS
PRIMARY COOLANT CIRCUITS
PUMPS
REACTIVITY
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
RESEARCH AND TEST REACTORS
SODIUM COOLED REACTORS