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

Zbigniew Koszela, Łukasz Sokołowski


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


nuclear safety analysis, LOCA, PWR

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