Due to the new security system that the operator intervention is assumed to occur in 20 minutes is not acceptable, for the current M310 type nuclear power plant. The loss of coolant accident with Intermediate breaks in primary loop is the only one of design basis accident which need operator action in 20 minutes. For certain size break, the consequences are very sensitive to the pump stop time. According to deterministic analysis that for a certain size break, if stop the pump in 20 minutes after accident, the peak cladding temperature will exceed the limit value of 1204°C. Therefore, it is necessary to add low-low pressurizer pressure in coincidence with high containment pressure signal to stop pump automatically on M310 type nuclear power plant.
Skip Nav Destination
2016 24th International Conference on Nuclear Engineering
June 26–30, 2016
Charlotte, North Carolina, USA
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5003-9
PROCEEDINGS PAPER
The Sensitivity of the Addition of Low-Low Pressurizer Pressure in Coincidence With High Containment Pressure on M310 Type Nuclear Power Plant
Qiu Yanfei
Qiu Yanfei
Suzhou Nuclear Power Research Institute of China, Shenzhen, China
Search for other works by this author on:
Qiu Yanfei
Suzhou Nuclear Power Research Institute of China, Shenzhen, China
Paper No:
ICONE24-60839, V003T09A064; 5 pages
Published Online:
October 25, 2016
Citation
Yanfei, Q. "The Sensitivity of the Addition of Low-Low Pressurizer Pressure in Coincidence With High Containment Pressure on M310 Type Nuclear Power Plant." Proceedings of the 2016 24th International Conference on Nuclear Engineering. Volume 3: Thermal-Hydraulics. Charlotte, North Carolina, USA. June 26–30, 2016. V003T09A064. ASME. https://doi.org/10.1115/ICONE24-60839
Download citation file:
13
Views
Related Proceedings Papers
Related Articles
Analyses of Feedwater Trip With SBO Sequence of VVER1000 Reactor
ASME J of Nuclear Rad Sci (October,2016)
Development of Safety Analysis Code TACOS and Application to Fuel Qualification Test Loop
ASME J of Nuclear Rad Sci (January,2017)
Assessment of the Code “PTCREEP” for IPHWR Pressure Tube Ballooning Study
J. Pressure Vessel Technol (February,2011)
Related Chapters
Insights and Results of the Shutdown PSA for a German SWR 69 Type Reactor (PSAM-0028)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Modeling of SAMG Operator Actions in Level 2 PSA (PSAM-0164)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
PSA Level 2 — NPP Ringhals 2 (PSAM-0156)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)