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THE ANALISYS OF TWO-PHASE LEVEL IN A PWR CORE DURING CONDITIONS OF SEVERELY REDUCED LIQUID MASS INVENTORY

DOI: 10.1615/ICHMT.1988.20thAHT.260
pages 317-333

Richard T. Lahey, Jr.
Center for Multiphase Research, Rensselaer Polytechnic Institute, Troy, NY 12180-3590, USA

Résumé

The recent incident at the Three Mile Island Nuclear Power Station, Unit #2 (TMI-2), has focused world wide attention on non-design-basis-accident (DBA) loss-of-coolant-accident (LOCA) phenomena. In particular, the need to be able to rapidly and accurately calculate the position of the two-phase level in a PWR core was clearly demonstrated at TMI-2. Indeed, as has been shown experimentally if the core is submerged in a two-phase mixture, the resultant boiling heat transfer is normally sufficient to remove the decay heat from the fuel rods, thus preventing overheating of the core.
While an evaluation of the position of the two-phase level is possible using existing digital computer codes (eg: RELAP/4, TRAC), this approach is quite costly and, in general, the information of interest is not available when needed by the operators experiencing a "small break" LOCA incident.
In this paper a closed form, analytical model for prediction of the two-phase level in a PWR core, during conditions of severely reduced liquid mass inventory, is derived. This model is based on drift-flux (4) techniques and assumes that thermodynamic equilibrium exists during the quasistatic process involved. It is easily shown that these assumptions are valid for many situations of interest.

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