Containment response to postulated ex-vessel core meltdown accidents in the fast flux test facility

Conference · Mon Jan 01 04:00:00 EST 1979

OSTI ID:6200646

Pratt, W T; Gasser, R D; Marchese, A R

An evaluation of the containment response to ex-vessel core meltdown accidents in the Fast Flux Test Facility is discussed. Particular attention is given to the possibility of the molten fuel/steel mixture remaining in a molten form after entering the reactor cavity. For the molten pool configurations considered, significant H/sub 2/ concentrations were reached in the RCB before the reactor cavity floor was penetrated. The melt-front impacts the pressure transient, not through gas generation, but through penetration into the unlined subcavity resulting in an enhanced sodium-concrete reaction. Hence, if H/sub 2/ recombination is neglected, then steps would have to be taken to decrease the H/sub 2/ concentration before the RCB design ressure is reached.

• Research Organization: Brookhaven National Lab., Upton, NY (USA); Nuclear Regulatory Commission, Washington, DC (USA). Div. of Project Management
• DOE Contract Number: EY-76-C-02-0016
• OSTI ID: 6200646
• Report Number(s): BNL-NUREG-25522; CONF-790816-1
• 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
BUILDING MATERIALS
CONCRETES
CONTAINMENT
CONTAINMENT BUILDINGS
CORIUM
EPITHERMAL REACTORS
FAST REACTORS
FFTF REACTOR
LIQUID METAL COOLED REACTORS
MATERIALS
MELTDOWN
PERFORMANCE
PRESSURE GRADIENTS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SAFETY
SODIUM COOLED REACTORS
STRESSES
TEST REACTORS
THERMAL STRESSES