Analysis of SBO ATWS for Maanshan PWR
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Che-Hao Chen
Abstract
Station blackout anticipated transient without scram (SBO ATWS) is considered as loss of off-site and on-site power but no credit for automatic reactor trip. SBO ATWS causes reactor coolant pump (RCP) trip, loss of all main feedwater pumps and turbine trip, then the reactor coolant system (RCS) pressure rises rapidly due to loss of heat removal paths. The ASME Code Level C service limit criteria of 22.06 MPa (3200 psig) is assumed to be an unacceptable plant condition in SECY-83-293. The simulation is performed by TRACE which is a thermal-hydraulic code developed by U.S. NRC. Three different AFW flows are modeled to ensure the pressures will not be beyond the criteria. RCP seal-leakage is concerned as a SBLOCA due to loss of RCP seal-cooling. Four possible leakage flows are modeled to examine the reactor core water level and temperature variation.
Kurzfassung
Ein vollständiger Ausfall der ungesicherten Wechselstromversorgung (Station Blackout, SBO) in Verbindung mit einer zu erwartenden Transiente ohne Reaktorschnellabschaltung (ATWS) wird betrachtet. SBO ATWS verursacht den Ausfall der Kühlmittelpumpen, der Hauptspeisewasserpumpen und einen Turbinenschnellschluss. Durch den Ausfall der Wärmeabfuhrpfade steigt der Druck im Reaktorkühlsystem schnell an. Der ASME Code Level C Druckgrenzwert von 22.06 MPa ist ein unzulässiger Anlagenzustand nach SECY-83-293. Die Simulation wurde mit Hilfe des Thermohydraulik-Codes TRACE durchgeführt. Drei verschiedene Ströme des Hilfsspeisewassersystems werden modelliert, um sicherzustellen, dass die Druckwerte nicht jenseits dieses Kriteriums liegen. Die Dichtungsleckage an den Kühlmittel pumpen wird als Kühlmittelverluststörfall mit kleinem Leck angenommen. Vier mögliche Leckageströme wurden modelliert, um den Wasserstand im Reaktorkern und die Temperaturschwankungen zu untersuchen.
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© 2015, Carl Hanser Verlag, München
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Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Theoretical study of steam condensation induced water hammer phenomena in horizontal pipelines
- Estimation of experimental uncertainty for physical measurements based on the start-up data of the latest VVER-1000 units
- Analysis of SBO ATWS for Maanshan PWR
- Subchannel analysis of Al2O3 nanofluid as a coolant in VMHWR
- The neutronic calculations for some fluids, libraries and structural materials in a hybrid reactor system
- Design and implementation progress of multi-purpose simulator for nuclear research reactor using LabVIEW
- International assessment of application of the Code of Conduct on the Safety of Research Reactors
- 15 MeV proton irradiation effects on Bi-based high temperature superconductors
- Estimation of radiation damage of iron by a reactor gamma spectrum
- Measuring U concentration in solution product of UF6 hydrolysis using a gamma ray densitometer
- Sensitivity analysis of parameters important to nuclear criticality safety of Castor X/28F spent nuclear fuel cask
- Application of UN method to neutron transport equation in slab geometry using HG phase function
- Preparation of human resources for future nuclear energy using FBNR as the instrument of learning
- Technical Note
- Interface network groups
