Sensitivity Study of the DEG LBLOCA Transient on the Counter-Current Flow Limitation by Using TRACE (NUREG/IA-0437)

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Publication Information

Manuscript Completed: September 2013
Date Published: February 2014

Prepared by:
Jong-Rong Wang, Che-Hao Chen*, 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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Abstract

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 a sensitivity study on the counter-current flow limitation (CCFL) model. The CCFL model plays a key role in any large break loss of coolant accident (LBLOCA) analysis since it affects the calculated discharge flow, reflooding time and peak cladding temperature (PCT). In this report, a sensitivity study on the CCFL model is performed, by modeling LBLOCA occurring at the Chinshan NPP. The scenario assumes 102% power and 75% core flow, with a double-ended guillotine (DEG) break on the recirculation loop, which is the most limiting LBLOCA for a BWR/4 reactor. Two break locations, i.e. on the suction and the discharge side of a recirculation pump, are evaluated, with high pressure core injecting (HPCI) and low pressure core spraying (LPCS) available whereas low pressure core injecting (LPCI) failed. The TRACE code is used for the analysis. The Chinshan TRACE model was benchmarked against steady-state and transient data contained in the plant FSAR report, as well as start-up data and the transient results using the RETRAN code. The thermal hydraulic phenomena in the lower plenum area and the jet pumps are also analyzed.