Containment response to postulated core meltdown accidents in the Fast Flux Test Facility

Technical Report · Tue Aug 01 04:00:00 EDT 1978

OSTI ID:6669088

Gasser, R D; Pratt, W T

An assessment of the containment margin available in the Fast Flux Test Facility to mitigate the consequences of a postulated failure on in-vessel post-accident heat removal following a hypothetical core-disruptive accident (HCDA) has been completed. Two general meltdown configurations (termed in-vessel and ex-vessel) have been considered, and it is concluded that: (1) For the in-vessel meltdown scenario, the quantity of sodium assumed ejected from the vessel into the cavity during the HCDA is crucial to the subsequent reactor containment building (RCB) pressurization. If the reactor cavity liner is also assumed to fail, then the containment integrity could be challenged within 10 hours. (2) For the ex-vessel meltdown scenario, the collapse of the reactor cavity floor, which allows sodium to contact the unlined subcavity, is crucial to the subsequent RCB pressurization. Predictions of cavity floor penetration vary between 2 to 60 hours depending on the initial assumptions made; however, after floor collapse occurs overpressurization of the RCB occurs rapidly within several hours.

• Research Organization: Brookhaven National Lab., Upton, NY (USA)
• DOE Contract Number: EY-76-C-02-0016
• OSTI ID: 6669088
• Report Number(s): BNL-NUREG-24141-R
• Country of Publication: United States
• Language: English

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
CONTAINMENT
EPITHERMAL REACTORS
FAST REACTORS
FFTF REACTOR
LIQUID METAL COOLED REACTORS
MELTDOWN
PRESSURE GRADIENTS
REACTOR ACCIDENTS
REACTOR CORE DISRUPTION
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SODIUM COOLED REACTORS
TEST REACTORS