Severe accident source term characteristics for selected Peach Bottom sequences predicted by the MELCOR Code

Technical Report · Wed Sep 01 00:00:00 EDT 1993

DOI:https://doi.org/10.2172/10182846· OSTI ID:10182846

Carbajo, J J ^[1]

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1. Oak Ridge National Lab., TN (United States)

The purpose of this report is to compare in-containment source terms developed for NUREG-1159, which used the Source Term Code Package (STCP), with those generated by MELCOR to identify significant differences. For this comparison, two short-term depressurized station blackout sequences (with a dry cavity and with a flooded cavity) and a Loss-of-Coolant Accident (LOCA) concurrent with complete loss of the Emergency Core Cooling System (ECCS) were analyzed for the Peach Bottom Atomic Power Station (a BWR-4 with a Mark I containment). The results indicate that for the sequences analyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the MELCOR/CORBH Package predicts significantly longer times for vessel failure and reduced energy of the released material for the station blackout sequences (when compared to the STCP results). MELCOR also calculated smaller releases into the environment than STCP for the station blackout sequences.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution; Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 10182846
• Report Number(s): NUREG/CR-5942; ORNL/TM-12229; ON: TI93041291; TRN: 93:021903
• Resource Relation: Other Information: DN: 1 copy to Berta Perez (ESTSC); PBD: Sep 1993
• 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-2 REACTOR
LOSS OF COOLANT
SOURCE TERMS
PEACH BOTTOM-3 REACTOR
BLACKOUTS
M CODES
CONTAINMENT
COMPUTER CALCULATIONS
CONTAINMENT SYSTEMS
S CODES
ECCS
HEAT TRANSFER
HYDRAULICS
REACTOR SAFETY
220900
210100
990200
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
BOILING WATER COOLED
MATHEMATICS AND COMPUTERS