Simulations of the recent LaSalle-2 incident with BNL plant analyzer
Conference
·
· Transactions of the American Nuclear Society; (USA)
OSTI ID:5619156
- Brookhaven National Lab., Upton, NY (USA)
On March 9, 1988, an instrument maintenance technician at LaSalle Unit 2, while performing a functional test on a differential pressure switch, caused both recirculation pumps to trip off due to a valving error. Because of the large and rapid power reduction, feedwater heater high level alarms caused a partial isolation of feedwater heaters, resulting in a reduction of 70{degree}F in feedwater temperature. Approximately 5 min into the event, the local power range monitor's up- and downscale alarms began annunciating, and the average power range monitors (APRMs) were observed to be oscillating with an {approx} 2.3-s period. Realizing the unit's unfavorable location on the power/flow map, the operating staff was preparing to scram the reactor manually, when an automatic scram occurred on a high-flux trip (118% trip on APROM). Prior to the scram, the operators attempted to remedy the situation by trying to restart the recirculation pumps, but failed. The diverging power oscillation observed in this incident is of concern to the US Nuclear Regulatory Commission (NRC) because of the implication that the reactor might have been unstable at the time of the incident. The Brookhaven National Laboratory plant analyzer has been used to simulate the recent LaSalle-2 power oscillation incident. By driving the reactor into the unstable region of the power/flow map, the authors were able to reproduce the diverging power oscillation as observed in the LaSalle-2 event. The conclusions drawn from the present work are outlined.
- OSTI ID:
- 5619156
- 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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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BNL
BWR TYPE REACTORS
COMPUTERIZED SIMULATION
CONTROL SYSTEMS
COOLING SYSTEMS
ELECTRICAL EQUIPMENT
ENERGY SYSTEMS
EQUIPMENT
ERRORS
FEEDWATER
HEAT EXCHANGERS
HUMAN FACTORS
HYDROGEN COMPOUNDS
LA SALLE COUNTY-2 REACTOR
MAINTENANCE
MODERATORS
NATIONAL ORGANIZATIONS
OXYGEN COMPOUNDS
PERFORMANCE TESTING
PERSONNEL
POWER DENSITY
POWER REACTORS
POWER-COOLING-MISMATCH ACCIDENTS
PRIMARY COOLANT CIRCUITS
PUMPS
RCIC SYSTEMS
REACTIVITY
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CONTROL SYSTEMS
REACTOR COOLING SYSTEMS
REACTOR CORES
REACTOR INSTRUMENTATION
REACTOR MAINTENANCE
REACTOR MONITORING SYSTEMS
REACTOR SHUTDOWN
REACTOR STABILITY
REACTORS
SCRAM
SENSITIVITY ANALYSIS
SHUTDOWN
SIMULATION
STABILITY
SWITCHES
TESTING
TRANSIENTS
US AEC
US DOE
US ERDA
US NRC
US ORGANIZATIONS
WATER
WATER COOLED REACTORS
WATER MODERATED REACTORS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BNL
BWR TYPE REACTORS
COMPUTERIZED SIMULATION
CONTROL SYSTEMS
COOLING SYSTEMS
ELECTRICAL EQUIPMENT
ENERGY SYSTEMS
EQUIPMENT
ERRORS
FEEDWATER
HEAT EXCHANGERS
HUMAN FACTORS
HYDROGEN COMPOUNDS
LA SALLE COUNTY-2 REACTOR
MAINTENANCE
MODERATORS
NATIONAL ORGANIZATIONS
OXYGEN COMPOUNDS
PERFORMANCE TESTING
PERSONNEL
POWER DENSITY
POWER REACTORS
POWER-COOLING-MISMATCH ACCIDENTS
PRIMARY COOLANT CIRCUITS
PUMPS
RCIC SYSTEMS
REACTIVITY
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CONTROL SYSTEMS
REACTOR COOLING SYSTEMS
REACTOR CORES
REACTOR INSTRUMENTATION
REACTOR MAINTENANCE
REACTOR MONITORING SYSTEMS
REACTOR SHUTDOWN
REACTOR STABILITY
REACTORS
SCRAM
SENSITIVITY ANALYSIS
SHUTDOWN
SIMULATION
STABILITY
SWITCHES
TESTING
TRANSIENTS
US AEC
US DOE
US ERDA
US NRC
US ORGANIZATIONS
WATER
WATER COOLED REACTORS
WATER MODERATED REACTORS