Fuel rod thermo-mechanic analysis of Kuosheng NPP in MSIVC transient with RELAP5 MOD3.3/Fraptran/Python code in SNAP interface

Journal Article · Fri Jul 01 04:00:00 EDT 2016 · Transactions of the American Nuclear Society

OSTI ID:23042873

Chang, H. C.; Shih, C.; Wang, J. R. ^[1]; Chiang, S. C.; Yu, T. Y. ^[2]; Lee, I. C. ^[3]; Hsu, W. S. ^[4]

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1. Institute of Nuclear Engineering and Science, National Tsing-Hua University, 101 Section 2, Kuang Fu Rd., HsinChu, Taiwan (China)
2. Department of Nuclear Safety, Taiwan Power Company, 242, Section 3, Roosevelt Rd., Zhongzheng District, Taipei, Taiwan (China)
3. Center for Energy and Environmental Research, National Tsing Hua University (China)
4. Nuclear Science and Technology Development Center, National Tsing Hua University (China)

Kuosheng NPP (Nuclear Power Plant) is located on the northern coast of Taiwan. Its nuclear steam supply system (NSSS) is a type of boiling water reactor (BWR/6) designed and built by General Electric on a twin unit concept. Each unit includes two loops of recirculation piping and four main steam lines, with the thermal rated power of 2894 MWt. Unit 1 started SPU (Stretch Power Uprate) from Cycle 24 and Unit 2 started SPU from Cycle 23. The operating power is 104.7% of the OLTP (Original Licensed Thermal Power), which is 3030 MWt now. In order to estimate the safety of Kuosheng NPP, the methodology of Kuosheng NPP SPU safety analysis model was developed. This methodology is shown in Fig. 1. There are three main steps considered in this methodology. The first step is the development of the Kuosheng NPP SPU RELAP5 model. The next step is the main steam line isolation valve closure (MSIVC) analysis of Kuosheng NPP SPU RELAP5 model in order to confirm the maximum pressure of vessel below the acceptance limit of 9.58 MPa. The final step is the fuel rods integrity analysis of FRAPTRAN under the above conditions. From FRAPTRAN analysis results, we find that the peak value of cladding hoop strain is smaller than 0.01, the acceptance limit; furthermore, the fuel enthalpy values are smaller than 150 cal/g. These two results indicate that the fuel rods are still in safe condition after turbine trip accident. (authors)

• OSTI ID: 23042873
• Journal Information: Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Vol. 115; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BWR TYPE REACTORS
CLADDING
CONTAINERS
ENTHALPY
FUEL RODS
MECHANICS
NUCLEAR POWER PLANTS
PYTHON
REACTOR DESIGN
SAFETY ANALYSIS
STEAM
STEAM LINES
TRANSIENTS
TURBINES
VALVES