Analysis of the turbine trip test performed at the Chin-Shan Power Station Unit 1 with plant analyzer
Conference
·
· Transactions of the American Nuclear Society; (United States)
OSTI ID:6808008
In order to improve nuclear power plant safety, plant analyzer technology has been transferred to Taiwan Power Company (TPC) from Brookhaven National Laboratory (BNL). The BNL boiling water reactor (BWR) plant analyzer program was converted and modified for CHIN-SHAN power station with ADSIM simulation language. The plant analyzer incorporates the special purpose peripheral processor AD-100 of Applied Dynamic International Company and detailed mathematical models. With the fast computing features of the AD-100, the plant analyzer can carry out accurate and fast simulations of nuclear power pant transients. In order to qualify this program, simulation results are benchmarked against power test data. During a power test of CHIN-SHAN power station unit 1 of TPC, a 1,775-MW(thermal) BWR/4, a turbine trip test was performed. This test was performed from a steady-state condition of 83% rated core power and 80% rated core flow. It was initiated by manual trip of the turbine. This turbine trip transient is simulated with the plant analyzer and compared with plant test data. The simulation results show that the plant analyzer catches the trends of this transient very well.
- OSTI ID:
- 6808008
- Report Number(s):
- CONF-901101--
- Conference Information:
- Journal Name: Transactions of the American Nuclear Society; (United States) Journal Volume: 62
- 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
ACCURACY
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BOILING
BWR TYPE REACTORS
BYPASSES
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COMPUTER CODES
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CONTAINERS
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ELECTRIC UTILITIES
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ENRICHED URANIUM REACTORS
FILM BOILING
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
KINETICS
MECHANICS
NATIONAL ORGANIZATIONS
PHASE TRANSFORMATIONS
PHYSICAL PROPERTIES
POWER DENSITY
POWER REACTORS
PRESSURE VESSELS
PRIMARY COOLANT CIRCUITS
PUBLIC UTILITIES
REACTIVITY
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR KINETICS
REACTOR MONITORING SYSTEMS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
REACTOR SHUTDOWN
REACTORS
SAFETY
SCRAM
SHUTDOWN
SIMULATION
THERMAL CONDUCTIVITY
THERMAL REACTORS
THERMODYNAMIC PROPERTIES
TIME DEPENDENCE
TRANSIENTS
TWO-PHASE FLOW
US AEC
US DOE
US ERDA
US ORGANIZATIONS
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
ACCURACY
BNL
BOILING
BWR TYPE REACTORS
BYPASSES
CHINSHAN-1 REACTOR
COMPUTER CODES
COMPUTERIZED SIMULATION
CONTAINERS
COOLING SYSTEMS
EFFICIENCY
ELECTRIC UTILITIES
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
FILM BOILING
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
KINETICS
MECHANICS
NATIONAL ORGANIZATIONS
PHASE TRANSFORMATIONS
PHYSICAL PROPERTIES
POWER DENSITY
POWER REACTORS
PRESSURE VESSELS
PRIMARY COOLANT CIRCUITS
PUBLIC UTILITIES
REACTIVITY
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR KINETICS
REACTOR MONITORING SYSTEMS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
REACTOR SHUTDOWN
REACTORS
SAFETY
SCRAM
SHUTDOWN
SIMULATION
THERMAL CONDUCTIVITY
THERMAL REACTORS
THERMODYNAMIC PROPERTIES
TIME DEPENDENCE
TRANSIENTS
TWO-PHASE FLOW
US AEC
US DOE
US ERDA
US ORGANIZATIONS
WATER COOLED REACTORS
WATER MODERATED REACTORS