Containment response to postulated core meltdown accidents in the fast flux test facility
An assessment is made of the containment margin available in the Fast Flux Test Facility to mitigate the consequences of a postulated failure of in-vessel post-accident heat removal following a hypothetical core disruptive accident. The consequences of a number of assumed meltdown configurations (both in-vessel and ex-vessel) are assessed using the CACECO (CAvty, CEll, COntainment) containment analysis computer code together with currently available melt front penetration models. The sensitivity of the accident scenarios to a number of crucial assumptions is established by scoping studies. It is concluded from both the in-vessel and exvessel analyses that sodium vapor combustion is a major source of reactor containment building (RCB) pressurization. The conditions (a combination of sodium-concrete reaction, pool size, and decay heat level) that most rapidly bring the sodium to boiling, together with those that enhance mass transfer of sodium vapor to the RCB, are the ones that most significantly affect the pressure response.
- Research Organization:
- Brookhaven National Lab., Upton, NY
- OSTI ID:
- 5488575
- Journal Information:
- Nucl. Technol.; (United States), Vol. 47:2
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FFTF REACTOR
MELTDOWN
RHR SYSTEMS
AFTER-HEAT REMOVAL
BOILING
COMPUTER CALCULATIONS
CONTAINMENT
FAILURES
PRESSURE GRADIENTS
SODIUM
ACCIDENTS
ALKALI METALS
COOLING SYSTEMS
ELEMENTS
EPITHERMAL REACTORS
FAST REACTORS
LIQUID METAL COOLED REACTORS
METALS
PHASE TRANSFORMATIONS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SODIUM COOLED REACTORS
TEST REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
220600 - Nuclear Reactor Technology- Research
Test & Experimental Reactors