Thermal-hydraulic analysis of single and multiple steam generator tube ruptures in a typical 3-loop PWR

• Autorzy: Koszela Z. , Sokołowski Ł.
• Języki publikacji: EN

Abstrakty

EN

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 different 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.

Słowa kluczowe

EN

nuclear safety analysis; nuclear safety; LOCA; PWR; Pressurized Water Reactor

PL

analiza bezpieczeństwa jądrowego; bezpieczeństwo jądrowe; LOCA; PWR; reaktor wodny ciśnieniowy

• Wydawca: Institute of Heat Engineering, Warsaw University of Technology
• Czasopismo: Journal of Power Technologies
• Rocznik: 2015
• Tom: Vol. 95, nr 3
• Strony: 175--182
• Opis fizyczny: Bibliogr. 13 poz., rys., tab., wykr.

Twórcy

autor

Koszela Z.

• Inspecta Nuclear AB, Lindhagensterrassen 1, 112 18 Stockholm, Sweden

autor

Sokołowski Ł.

• Inspecta Nuclear AB, Lindhagensterrassen 1, 112 18 Stockholm, Sweden

Bibliografia

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