Assessment of the safety function for the anticipated transient without trip mitigation system actuation circuitry at Maanshan Nuclear Power Station
Journal Article
·
· Nuclear Technology; (USA)
OSTI ID:6499259
- National Tsing-Hua Univ., Dept. of Nuclear Engineering, Hsinchu 30043 (TW)
Due to the potential threat of reactor coolant system (RCS) overpressurization, loss-of-normal-feed-water (LONF) transients without reactor trip have received special attention in the analysis of pressurized water reactor (PWR) anticipated transients without trip (ATWT). The U.S. Nuclear Regulatory Commission requires every PWR to be equipped with an ATWT mitigation system actuation circuitry (AMSAC) so that the turbine will be tripped and auxiliary feedwater (AFW) added when an LONF transient occurs. An AMSAC design will be installed in both units of the Maanshan Nuclear Power Station (MNPS) to deal with ATWTs under LONF transient conditions. A best-estimate transient analysis performed with the RETRAN-02/MOD3 code is used to assess the safety function of the actuation circuitry designed for MNPS. Analytical results show that the peak RCS pressure will not exceed the 22.16-MPa safety limit if the moderator temperature coefficient is sufficiently negative and the actuation circuitry functions normally. Effects of the moderator temperature coefficient, the Doppler coefficient, pressurizer power-operated relief valves, effective time of the AFW system, the steam dump system, and the automatic control rod system are discussed.
- OSTI ID:
- 6499259
- Journal Information:
- Nuclear Technology; (USA), Journal Name: Nuclear Technology; (USA) Vol. 90:1; ISSN NUTYB; ISSN 0029-5450
- Country of Publication:
- United States
- Language:
- English
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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
220400 -- Nuclear Reactor Technology-- Control Systems
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
AUTOMATION
CONTROL EQUIPMENT
CONTROL SYSTEMS
DESIGN
EQUIPMENT
FLOW REGULATORS
LOSS OF COOLANT
NATIONAL ORGANIZATIONS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY EXPERIMENTS
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TESTING
US NRC
US ORGANIZATIONS
VALVES
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220400 -- Nuclear Reactor Technology-- Control Systems
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
AUTOMATION
CONTROL EQUIPMENT
CONTROL SYSTEMS
DESIGN
EQUIPMENT
FLOW REGULATORS
LOSS OF COOLANT
NATIONAL ORGANIZATIONS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY EXPERIMENTS
REACTORS
TESTING
US NRC
US ORGANIZATIONS
VALVES
WATER COOLED REACTORS
WATER MODERATED REACTORS