Containment response to postulated core meltdown accidents in the fast flux test facility
Journal Article
·
· Nucl. Technol.; (United States)
OSTI ID:5488575
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), Journal Name: Nucl. Technol.; (United States) Vol. 47:2; ISSN NUTYB
- Country of Publication:
- United States
- Language:
- English
Similar Records
Containment response to postulated core meltdown accidents in the Fast Flux Test Facility
Containment response to postulated ex-vessel core meltdown accidents in the fast flux test facility
Containment transient analysis for postulated accident consequences assessment. [LMFBR]
Technical Report
·
Tue Aug 01 00:00:00 EDT 1978
·
OSTI ID:6669088
Containment response to postulated ex-vessel core meltdown accidents in the fast flux test facility
Conference
·
Sun Dec 31 23:00:00 EST 1978
·
OSTI ID:6200646
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Conference
·
Tue Mar 30 23:00:00 EST 1976
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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
ALKALI METALS
BOILING
COMPUTER CALCULATIONS
CONTAINMENT
COOLING SYSTEMS
ELEMENTS
EPITHERMAL REACTORS
FAILURES
FAST REACTORS
FFTF REACTOR
LIQUID METAL COOLED REACTORS
MELTDOWN
METALS
PHASE TRANSFORMATIONS
PRESSURE GRADIENTS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
RHR SYSTEMS
SODIUM
SODIUM COOLED REACTORS
TEST REACTORS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
AFTER-HEAT REMOVAL
ALKALI METALS
BOILING
COMPUTER CALCULATIONS
CONTAINMENT
COOLING SYSTEMS
ELEMENTS
EPITHERMAL REACTORS
FAILURES
FAST REACTORS
FFTF REACTOR
LIQUID METAL COOLED REACTORS
MELTDOWN
METALS
PHASE TRANSFORMATIONS
PRESSURE GRADIENTS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
RHR SYSTEMS
SODIUM
SODIUM COOLED REACTORS
TEST REACTORS