Thermal-Hydraulic Analysis of Single and Multiple Steam Generator Tube Ruptures in a Typical 3-Loop PWR

  • Zbigniew Koszela Inspecta Nuclear AB
  • Łukasz Sokołowski Inspecta Nuclear AB

Abstract

The response of the full-scale three-loop Pressurized Water Reactor (PWR) RELAP5 computational model on SteamGenerator Break Rupture (SGBR) was investigated in this paper. This model was analyzed in terms of its applicabilityand performance regarding the research task conducted by Warsaw University of Technology and the National Centerfor Research and Development inWarsaw, Poland. In the paper break sizes corresponding to one, three and six rupturedtubes (which conform to a Loss-of-Coolant event break size area of 0.02%, 0.054 and 0.11%) were studied at threedierent locations (at the top of the hot-leg side tubesheet, U-bend and at the top of the cold-leg side tubesheet). Thereactor at issue was a three-loop PWR of Westinghouse design with thermal output of 2775 MWt.

Author Biographies

Zbigniew Koszela, Inspecta Nuclear AB
Dr Zbigniew Koszela is a graduate of Warsaw University of Technology, Poland, and has technical background in safety analyses of Light Water Reactors.
Łukasz Sokołowski, Inspecta Nuclear AB
Lukasz Sokolowski has graduated in Nuclear Energy Engineering from Royal Institute of Technology in Stockholm, Sweden. His technical background is thermal-hydraulics of Light Water Reactors and multiphase flows.
Published
2015-07-04
How to Cite
KOSZELA, Zbigniew; SOKOŁOWSKI, Łukasz. Thermal-Hydraulic Analysis of Single and Multiple Steam Generator Tube Ruptures in a Typical 3-Loop PWR. Journal of Power Technologies, [S.l.], v. 95, n. 3, p. 175--182, july 2015. ISSN 2083-4195. Available at: <https://papers.itc.pw.edu.pl/index.php/JPT/article/view/584>. Date accessed: 19 apr. 2024.
Citation Formats
Issue
Vol 95 No 3 (2015)
Section
Nuclear Power

Keywords

nuclear safety analysis, LOCA, PWR
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