Thermal-Hydraulic Analysis of Single and Multiple Steam Generator Tube Ruptures in a Typical 3-Loop PWR
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.
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: 23 nov. 2024.
Issue
Section
Nuclear Power
Keywords
nuclear safety analysis, LOCA, PWR
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