Safety analysis of bleed-and-feed operations for Maanshan power plant

Conference · Tue Jan 01 04:00:00 EST 1991 · Transactions of the American Nuclear Society; (United States)

OSTI ID:5977965

Kao, Lainsu ^[1]; Fan, Youn; Pei, Baushei; Yu, Geping ^[2]

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1. Inst. of Nuclear Energy Research, Lung-Tan (Taiwan)
2. National Tsing Hua Univ., Hsinchu (Taiwan)

Thermal-hydraulic calculations using RELAP5/MOD2 are performed to simulate the responses of the Maanshan pressurized water reactor (PWR) power plant when the secondary heat sink is lost. The loss of heat sink transient of primary interest in this study is the loss of off-site power coincident with a loss-of-all-feedwater event. This transient is selected because it is one of the risk-dominant sequences leading to core damage for the Maanshan PWR power plant. The initiating event for the transient is assumed to be loss of offsite power, which trips the turbine, the reactor coolant pumps, and the main feedwater pumps and generates a signal to scram the reactor. Neither the motor-driven nor the turbine-driven auxiliary feedwater system supply make up water to the steam generator secondary on demand. According to the emergency operation procedures (EOPs), the operator should make every effort to restore the function of the auxiliary feedwater systems. If such efforts cannot succeed in the recovery of the secondary heat sink before the secondary inventory dryout indicated by 6% readings of the wide range water-level measurements, bleed-and-feed (B and F) operations should be executed in time to prevent damage to the integrity of the reactor core. Once the operator decides to take the B and F action, the EOP requires the operator to open at least two out of three pressurizer power-operated relief valves to depressurize the primary system and at the same time, manually, introduce HHSI with at least minimum flow capacity to cool down the reactor core. The major purpose of this study is to identify the effectiveness of the B and F operation in mitigating the consequences of the event.

• OSTI ID: 5977965
• Report Number(s): CONF-910603--
• Conference Information: Journal Name: Transactions of the American Nuclear Society; (United States) Journal Volume: 63
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
AC SYSTEMS
ACCIDENTS
COMPUTER CODES
COMPUTERIZED SIMULATION
CONTROL EQUIPMENT
COOLING SYSTEMS
D CODES
DEPRESSURIZATION
EMERGENCY PLANS
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
EQUIPMENT
FEEDWATER
FLOW RATE
FLOW REGULATORS
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
HYDROGEN COMPOUNDS
LOSS OF COOLANT
MAANSHAN-1 REACTOR
MECHANICS
MOTORS
OPERATION
OUTAGES
OXYGEN COMPOUNDS
PERSONNEL
POWER REACTORS
POWER SYSTEMS
PRESSURIZERS
PRIMARY COOLANT CIRCUITS
PUMPS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR CORE DISRUPTION
REACTOR CORES
REACTOR OPERATION
REACTOR OPERATORS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
REACTOR SHUTDOWN
REACTORS
RELIEF VALVES
RHR SYSTEMS
SAFETY
SCRAM
SECONDARY COOLANT CIRCUITS
SHUTDOWN
SIMULATION
THERMAL REACTORS
TIME DEPENDENCE
VALVES
WATER
WATER COOLED REACTORS
WATER MODERATED REACTORS