Analyses of corium spreading in Mark I containment geometry

Conference · Tue Dec 31 00:00:00 EST 1991

OSTI ID:10107094

Sienicki, J J; Chu, C C; Farmer, M T

An assessment of melt spreading in the Mark I system has been carried out using the MELTSPREAD-1 computer code together with supporting analyses. Application of MELTSPREAD-1 confirms the calculation of shell survival in a wet containment for the most probable melt release conditions from NUREG/CR-5423. According to MELTSPREAD-1, a dry containment also may not be threatened by melt spreading. This reflects the heat losses undergone by the melt in the process of spreading to the shell conservatively neglected in NUREG/CR-5423. However, there exist parameter ranges outside the most probable set where shell failure may be calculated. Accounting for the breakup and quenching of melt relocating through a deep layer of subcooled water also conservatively neglected in NUREG/CR-5423 can reduce the set of parameter variations for which containment failure is calculated in the wet case.

• Research Organization: Argonne National Lab., IL (United States)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: W-31109-ENG-38
• OSTI ID: 10107094
• Report Number(s): ANL/CP-74781; CONF-911079-13; ON: DE92004140
• Resource Relation: Conference: 19. Nuclear Regulatory Commission (NRC) water reactor safety information meeting,Bethesda, MD (United States),28-30 Oct 1991; Other Information: PBD: [1991]
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BWR TYPE REACTORS
MELTDOWN
CORIUM
CONTAINMENT
M CODES
CONTAINMENT SHELLS
HEAT LOSSES
HEAT TRANSFER
HYDRAULICS
REACTOR SAFETY
SENSITIVITY ANALYSIS
220900
210100
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
BOILING WATER COOLED