The Establishment and Assessment of Chinshan (BWR/4) Nuclear Power Plant TRACE/SNAP Model (NUREG/IA-0445)

On this page:

• Publication Information
• Abstract

Download complete document

• NUREG/IA-0445 (PDF - 5.32 MB)

Publication Information

Manuscript Completed: April 2014
Date Published: July 2014

Prepared by:
Jong-Rong Wang, Hao-Tzu Lin, Hsiung-Chih Chen*, 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

Availability Notice

Abstract

Chinshan Nuclear Power Plant (NPP) is the first NPP in Taiwan which is of the BWR/4 type plant. This research focuses on the development of the Chinshan NPP TRACE/SNAP model. In order to check the system response of the Chinshan NPP TRACE/SNAP model, this study uses the analysis results of Final Safety Analysis Report (FSAR) to assess the Chinshan NPP TRACE/SNAP model. The increase in reactor pressure transients including turbine trip and main steam isolation valves closure were selected to validate the Chinshan NPP TRACE/SNAP model. The trends of TRACE analysis results were consistent with the FSAR data. It indicates that there is a credible fidelity in the Chinshan NPP TRACE/SNAP model. In addition, this research also investigates the application of the Chinshan NPP TRACE/SNAP model for the core shroud leakage. The core shroud leakage is one of current issues of concern of the U.S. NRC and other BWR/4 NPP owners. This research utilizes the Chinshan NPP TRACE/SNAP model to perform the core shroud leakage transients for Chinshan NPP safety analysis. The TRACE analysis results show that the simple core shroud leakage transient does not influence the Chinshan NPP fuel temperature to cause any core damage. However, the core shroud leakage combined with station blackout (SBO) or steamline break (loss of coolant accident, LOCA) transient would cause the cladding temperature to reach higher than 1088 K and affected the Chinshan NPP safety.