Uncertainty Analysis of Main Steam Line Break Accident for Maanshan PWR with RELAP5/DAKOTA (NUREG/IA-0528)

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Publication Information

Manuscript Completed: September 2021
Date Published: February 2022

Prepared by:
Chunkuan Shih, Jong-Rong Wang, Chih-Chia Chiang, Yuh-Ming Ferng, Shao-Wen Chen, and Tzu-Yao Yu*

National Tsing Hua University and Nuclear and New Energy 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)

Availability Notice


In our previous study, the RELAP5/SNAP model of Maanshan PWR nuclear power plant is established. This model was used to perform the analysis of Main Steam Line break (MSLB) inside-containment transient. The analysis results of RELAP5/SNAP are consistent with the FSAR data. In this study, the main purpose is to perform an uncertainty analysis for Maanshan MSLB by using RELAP5/SNAP model and DAKOTA code. Total 21 parameters which include initial power, accumulator volume, injection water temperature, injection flow, rod material thermal conductivity, discharge coefficient for break, slug flow drag, etc. are evaluated in this analysis. According to the uncertainty analysis results, discharge coefficient for break, slug flow drag, and annular-mist flow drag have larger effect in the calculation of break flow, and slug flow drag and annular-mist flow drag have larger effect in the calculation of void fraction.

Page Last Reviewed/Updated Wednesday, February 23, 2022