Maanshan T sub p S sub L B accident analysis

Conference · Fri Jan 01 04:00:00 EST 1988 · Transactions of the American Nuclear Society; (USA)

OSTI ID:5497243

Chen, C C; Wang, T K; Hsueh, J K

An T{sub p}S{sub L}B accident analysis for Taipower's Maanshan Unit 1 plant is reported. The plant is a 2775-MW(thermal) pressurized water reactor with large dry containment. Based on Maanhshan level-1 probabilistic risk assessment, the T{sub p}S{sub L}B sequence ranks first in accident frequency. The basic definition of T{sub p}S{sub L}B includes loss-of-off-site power T{sub p}, station blackout, loss of all alternating current power B, and the occurrence of a loss-of-coolant accident (LOCA) due to main coolant pump seal failure, i.e., seal LOCA S{sub L}. The results of a base case study are summarized first. These results are then compared with other studies by varying some key parameter values. All calculations were done with MAAP 3.0 computer code. In the base case study, seal LOCA was assumed to occur 3 h after the initiation of the vent. The LOCA break area was assumed to be 0.169 x 10{sup {minus}3} m{sup 2}. Leakage occurred in all three loops. Containment (CTMT) failure pressure was set at 120 psig. The containment break area was set at 0.003 m{sup 2}, attempting to keep the pressure roughly constant after containment failure. In the station blackout situation, the direct current power in the Maanshan plant would last 2 h. Accordingly, in the calculations we assumed the turbine-driven auxiliary feedwater and the power-operated relief valves would also be working for 2 h. The major event sequences, including times of core uncovery, vessel failure, and containment failure as well as the resulting fission product release fractions at different time spans after containment failure, are listed. Also listed are the results of other studies.

• OSTI ID: 5497243
• Report Number(s): CONF-881011--
• Conference Information: Journal Name: Transactions of the American Nuclear Society; (USA) Journal Volume: 57
• Country of Publication: United States
• Language: English

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220900* -- Nuclear Reactor Technology-- Reactor Safety
AC SYSTEMS
ACCIDENTS
ALKALI METALS
BLACKOUTS
BUILDING MATERIALS
CESIUM
CHEMICAL REACTIONS
COMBUSTION
COMPUTER CODES
COMPUTERIZED SIMULATION
CONCRETES
CONTAINERS
CONTAINMENT
CONTAINMENT SYSTEMS
CONTROL EQUIPMENT
COOLING SYSTEMS
CORIUM
ELEMENTS
ENERGY SYSTEMS
ENGINEERED SAFETY SYSTEMS
ENTRAINMENT
EQUIPMENT
FAILURES
FEEDWATER
FISSION PRODUCT RELEASE
FLOW REGULATORS
HYDROGEN
HYDROGEN COMPOUNDS
LOSS OF COOLANT
M CODES
MATERIALS
METALS
MOLTEN METAL-WATER REACTIONS
NONMETALS
OUTAGES
OXIDATION
OXYGEN COMPOUNDS
POWER SYSTEMS
PRESSURE VESSELS
PRIMARY COOLANT CIRCUITS
PUMPS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR CORE DISRUPTION
REACTOR SAFETY
RELIEF VALVES
RISK ASSESSMENT
SAFETY
SAFETY INJECTION
SEALS
SIMULATION
THERMOCHEMICAL PROCESSES
VALVES
WATER