Qinshan Phase III is the first commercial pressurized heavy water reactor (PHWR) NPP in China, and it uses CANDU-6 design developed by AECL. Based on plant design and operation experience, the event tree analysis model has been developed for both small break LOCA (SB-LOCA) and large break LOCA (LB-LOCA), which is an important aspect of operational Probabilistic Safety Assessment (PSA). Both SB-LOCA and LB-LOCA event tree analysis have been performed for Qinshan Phase III CANDU-6 PHWR NPP (TQNPC). And the event sequence development and plant damage status (PDS) were provided in the analysis. It reflects actual plant configuration and response under a certain event, and various break type and locations were also considered in the event tree analysis, e.g. Pressure Tube Rupture, Pressure Tube and Calandria Tube Rupture, Feeder Breaks, Pressurizer Relief/Steam Bleed Valves Fail Open, Liquid Relief Valves Fail Open, etc.
Skip Nav Destination
18th International Conference on Nuclear Engineering
May 17–21, 2010
Xi’an, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-4929-3
PROCEEDINGS PAPER
LOCA Event Tree Analysis for Qinshan Phase III CANDU-6 PHWR NPP
Limin Zheng
Limin Zheng
Shanghai Nuclear Engineering Research & Design Institute (SNERDI), Shanghai, China
Search for other works by this author on:
Limin Zheng
Shanghai Nuclear Engineering Research & Design Institute (SNERDI), Shanghai, China
Paper No:
ICONE18-29746, pp. 251-259; 9 pages
Published Online:
April 8, 2011
Citation
Zheng, L. "LOCA Event Tree Analysis for Qinshan Phase III CANDU-6 PHWR NPP." Proceedings of the 18th International Conference on Nuclear Engineering. 18th International Conference on Nuclear Engineering: Volume 1. Xi’an, China. May 17–21, 2010. pp. 251-259. ASME. https://doi.org/10.1115/ICONE18-29746
Download citation file:
18
Views
Related Proceedings Papers
Related Articles
Analysis of the Thermal Hydraulic Consequences Following Common Mode Pump Seizure in a Nuclear Power Plant
J. Pressure Vessel Technol (November,2002)
Decoupling Criteria for Multi-Connected Equipment
J. Pressure Vessel Technol (February,1998)
Current Status of Reactors Deployment and Small Modular Reactors Development in the World
ASME J of Nuclear Rad Sci (October,2020)
Related Chapters
Design of Indian Pressurized Heavy Water Reactors
Global Applications of the ASME Boiler & Pressure Vessel Code
Design of Indian Pressurized Heavy Water Reactor Components
Companion Guide to the ASME Boiler and Pressure Vessel Code, Volume 3, Third Edition
Modeling of SAMG Operator Actions in Level 2 PSA (PSAM-0164)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)