Contain calculations of debris conditions adjacent to the BWR Mark I drywell shell during the later phases of a severe accident

Conference · Fri Jan 01 00:00:00 EST 1988

OSTI ID:6659154

Hyman, C R

Best estimate CONTAIN calculations have recently been performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to predict the primary containment response during the later phases of an unmitigated low-pressure Short Term Station Blackout at the Peach Bottom Atomic Power Station. Debris pour conditions leaving the failed reactor vessel are taken from the results of best estimate BWRSAR analyses that are based upon an assumed metallic debris melting temperature of 2750/degree/F (1783 K) and an oxide debris melting temperature of 4350/degree/F (2672 K). Results of the CONTAIN analysis for the case without sprays indicate failure of the drywell seals due to the extremely hot atmospheric conditions extant in the drywell. The maximum calculated temperature of the debris adjacent to the drywell shell is less than the melting temperature of the shell, yet the sustained temperatures may be sufficient to induce primary containment pressure boundary failure by the mechanism of creep-rupture. It is also predicted that a significant portion of the reactor pedestal wall is ablated during the period of the calculation. Nevertheless, the calculated results are recognized to be influenced by large modeling uncertainties. Several deficiencies in the application of the CORCON module within the CONTAIN code to BWR severe accident sequences are identified and discussed. 5 refs., 9 figs., 4 tabs.,

• Research Organization: Oak Ridge National Lab., TN (USA)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 6659154
• Report Number(s): CONF-8810155-19; ON: DE89003179
• Resource Relation: Conference: 16. water reactor safety information meeting, Gaithersburg, MD, USA, 24 Oct 1988; Other Information: Portions of this document are illegible in microfiche products
• 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
PEACH BOTTOM-1 REACTOR
CONTAINMENT
PEACH BOTTOM-2 REACTOR
PEACH BOTTOM-3 REACTOR
BLACKOUTS
C CODES
CORIUM
HEAT TRANSFER
HYDRAULICS
REACTOR ACCIDENTS
REACTOR SAFETY
SEALS
ACCIDENTS
BWR TYPE REACTORS
COMPUTER CODES
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLUID MECHANICS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
HELIUM COOLED REACTORS
HTGR TYPE REACTORS
MECHANICS
POWER REACTORS
REACTORS
SAFETY
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
990220 - Computers
Computerized Models
& Computer Programs- (1987-1989)