The Alternate Mitigation Strategies Study of Chinshan BWR/4 by Using the LOCA and SBO Analysis of TRACE (NUREG/IA-0440)

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Publication Information

Manuscript Completed: September 2013
Date Published: March 2014

Prepared by:
Jong-Rong Wang, Ai-Ling Ho*, Hao-Tzu Lin, Chunkuan Shih*

Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.
1000, Wenhua Rd., Chiaan Village, Lungtan, Taoyuan, 325, Taiwan

*Institute of Nuclear Engineering and Science, National Tsing Hua University
101 Section 2, Kuang Fu Rd., HsinChu, Taiwan

K. Tien, NRC Project Manager

Division of Systems Analysis
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

Prepared as part of:
The Agreement on Research Participation and Technical Exchange
Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by:
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

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Chinshan nuclear power plant is the first NPP in Taiwan which is the BWR/4 plant. This research focuses on the development of the Chinshan NPP TRACE model and the LOCA combined SBO accident analysis. From the accident at the Japanese Fukusima NPP, an extreme event beyond the design basis is realized to be possible. The current mitigation strategies for the emergency core cooling systems (ECCSs) can be easily voided in the event of an extended station blackout (SBO), where all the onsite and offsite electrical power is failed. Although the electrical power of the critical control systems can be recovered by portable electrical generators, the electrical pumps are difficult to recover by any portable device. The only possible driving force of the pumps in SBO is the steam generated by residual heat. The current strategies in an extended SBO are mostly focused on low pressure injection, but the reactor water level will decrease sharply while the reactor pressure is reduced and that results in a higher PCT. In this report, the alternate mitigation strategies adopting the turbine driven pumps, the high pressure injection systems, are analyzed to maintain an “enough” water level before the reactor pressure is reduced. Three break sizes, 100%, 10% and 1%, on the recirculation suction line of Chinshan NPP which is the most serious LOCA in BWR/4 reactor are analyzed with three sensitivity studies: (1) the scram time, (2) the increase of RCIC injection flow rate, and (3) the earlier HPCI injection.

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