Shin-Kori Units 3&4 (SKN 3&4) are the first APR1400 (Advanced Power Reactor 1400 MWe) nuclear power plants being under construction in Korea. The capacity and size of main components in the APR1400 nuclear steam supply system have been increased from the OPR1000 (Optimized Power Reactor 1000 MWe) which is the reference plant in operation. Especially, the increased pipe size of the main steam and the economizer feedwater lines cause more severe excitations such as the nozzle thrust, jet impingement, and nozzle reactions when those lines are postulated to break at terminal ends or intermediate locations. In this paper, the major mechanical effects on the Reactor Coolant System (RCS) due to the branch line pipe breaks for SKN 3&4 are investigated; the support loads of the reactor vessel, steam generators, reactor coolant pumps are evaluated. In this investigation, the RCS is modeled using 3-dimensional lumped mass and beam elements, and the nonlinear time-history structural analyses are performed. The analysis results show that the RCS dynamic responses are governed by the main steam line breaks and feedwater economizer line breaks.
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18th International Conference on Nuclear Engineering
May 17–21, 2010
Xi’an, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-4933-0
PROCEEDINGS PAPER
Reactor Coolant System Branch Line Pipe Break Analysis for Shin-Kori Units 3 and 4
Young Jin Byun,
Young Jin Byun
Korea Power Engineering Company, Inc., Daejeon, Korea
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Young Sil Sul,
Young Sil Sul
Korea Power Engineering Company, Inc., Daejeon, Korea
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Kwang Won Lee,
Kwang Won Lee
Korea Power Engineering Company, Inc., Daejeon, Korea
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Jong Tae Seo
Jong Tae Seo
Korea Power Engineering Company, Inc., Daejeon, Korea
Search for other works by this author on:
Young Jin Byun
Korea Power Engineering Company, Inc., Daejeon, Korea
Young Sil Sul
Korea Power Engineering Company, Inc., Daejeon, Korea
Kwang Won Lee
Korea Power Engineering Company, Inc., Daejeon, Korea
Jong Tae Seo
Korea Power Engineering Company, Inc., Daejeon, Korea
Paper No:
ICONE18-30066, pp. 639-645; 7 pages
Published Online:
April 8, 2011
Citation
Byun, YJ, Sul, YS, Lee, KW, & Seo, JT. "Reactor Coolant System Branch Line Pipe Break Analysis for Shin-Kori Units 3 and 4." Proceedings of the 18th International Conference on Nuclear Engineering. 18th International Conference on Nuclear Engineering: Volume 5. Xi’an, China. May 17–21, 2010. pp. 639-645. ASME. https://doi.org/10.1115/ICONE18-30066
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