Fuel Rod Behavior and Uncertainty Analysis by FRAPTRAN/TRACE/DAKOTA Code in Maanshan LBLOCA (NUREG/IA-0471)

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Publication Information

Manuscript Completed: March 2016
Date Published: August 2016

Prepared by:
Chunkuan Shih, Jung-Hua Yang, Jong-Rong Wang, Shao-Wen Chen, Show-Chyuan Chiang*, Tzu-Yao Yu*

Education and Research Foundation
101 Section 2, Kuang Fu Rd., HsinChu, Taiwan

*Department of Nuclear Safety, Taiwan Power Company
242, Section 3, Roosevelt Rd., Zhongzheng District, Taipei, 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


In this study, the FRAPTRAN and TRACE code were used to evaluate the fuel rod transient behavior during a postulated LBLOCA in Maanshan (3-loops PWR) Nuclear Power Plant (NPP). There were three main steps in this research. The first step was the LBLOCA analysis for Maanshan NPP by TRACE code. The analysis results were benchmarked and compared with Maanshan FSAR data. In second step, the geometry data of the fuel rod and the results from TRACE analysis (e.g. fuel rod power, coolant pressure, heat transfer coefficient) were input into FRAPTRAN to analyze the reliability of fuel rod. Then, it used FRAPTRAN to calculate the response of a single fuel rod transient behavior during LBLOCA. FRAPTRAN can obtain the detail mechanical property of fuel rod (e.g. cladding temperature, hoop stress/strain, gap pressure, and oxide thickness of cladding). After all, uncertainty analysis was considered in this study. The several parameters of fuel rod, such as fabrication and boundary conditions, were quantized and sampled by the DAKOTA uncertainty code.

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