An analysis of molten-corium-induced failure of drain pipes in BWR Mark 2 containments

Conference · Tue Jan 01 00:00:00 EST 1991

OSTI ID:5681831

Taleyarkhan, R P ^[1]; Podowski, M Z ^[2]

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1. Oak Ridge National Lab., TN (USA)
2. Rensselaer Polytechnic Inst., Troy, NY (USA)

This study has focused on mechanistic simulation and analysis of potential failure modes for inpedestal drywell drain pipes in the Limerick boiling water reactor (BWR) Mark 2 containment. Physical phenomena related to surface tension breakdown, heatup, melting, ablation, crust formation and failure, and core material relocation into drain pipes with simultaneous melting of pipe walls were modeled and analyzed. The results of analysis have been used to assess the possibility of drain pipe failure and the resultant loss of pressure-suppression capability. Estimates have been made for the timing and amount of molten corium released to the wetwell. The study has revealed that significantly different melt progression sequences can result depending upon the failure characteristics of the frozen metallic crust which forms over the drain cover during the initial stages of debris pour. Another important result is that it can take several days for the molten fuel to ablate the frozen metallic debris layer -- if the frozen layer has cooled below 1100 K before fuel attack. 10 refs., 3 figs., 4 tabs.

• Research Organization: Oak Ridge National Lab., TN (USA)
• Sponsoring Organization: USDOE; USNRC; USDOE, Washington, DC (USA); Nuclear Regulatory Commission, Washington, DC (USA)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 5681831
• Report Number(s): CONF-910739-13; ON: DE91011365
• Resource Relation: Conference: ASME/AIChE/ANS national heat transfer conference, Minneapolis, MN (USA), 26-31 Jul 1991
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
36 MATERIALS SCIENCE
LIMERICK-2 REACTOR
REACTOR SAFETY EXPERIMENTS
PIPES
MOLTEN METAL-WATER REACTIONS
REACTOR CORE DISRUPTION
COMPUTERIZED SIMULATION
ABLATION
B CODES
CORIUM
FAILURE MODE ANALYSIS
HEAT TRANSFER
HYDRAULICS
MELTDOWN
PRESSURE VESSELS
REACTOR SAFETY
STEELS
ZIRCONIUM
ACCIDENTS
ALLOYS
BWR TYPE REACTORS
COMPUTER CODES
CONTAINERS
ELEMENTS
ENERGY TRANSFER
FLUID MECHANICS
IRON ALLOYS
IRON BASE ALLOYS
MECHANICS
METALS
POWER REACTORS
REACTOR ACCIDENTS
REACTORS
SAFETY
SIMULATION
SYSTEM FAILURE ANALYSIS
SYSTEMS ANALYSIS
TRANSITION ELEMENTS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
990200 - Mathematics & Computers
360105 - Metals & Alloys- Corrosion & Erosion