Abstract
The onset of CCFL determines the point at which the cooling liquid flow is partially limited into the reactor vessel meaning a less cooling potential of the core during LOCA accidents. It is of a special importance for the safety of PWR reactors during the reflux condensation mode following an SBLOCA accident. Because a limited number of experimental measurements of the water film depth during a supercritical flow exist, a numerical model of a PWR hot-leg pipe geometry using ANSYS CFX 16.0 was prepared at different scales/pipe diameters: 50,80,190, and 750 mm which corresponds to pipes diameters used in previous experimental investigations of CCFL at a hot-leg pipe geometry. 3D Simulations were conducted using a two-fluid Euler-Euler approach and a validated model of the interfacial drag coefficient. A correlation between the water film thickness within the horizontal part at three selected locations: X/D=0.84, 4.44,8.04 and the liquid inlet flow rate, pipe diameter, and location was extracted out of calculations. The correlation was then used as an input parameter for a simplified model to predict the supercritical/subcritical transition and derive a new correlation of the interfacial friction factor in the horizontal part of the hot-leg. This limit was experimentally measured in several previous investigations at above mentioned pipe diameters and is known as the onset of the hydraulic jump (for liquid flow rates below the onset of hysteresis limit), and the onset of countercurrent flow limitation (for liquid flow rates above the onset of hysteresis limit). New correlations can be implemented in 1D system codes to predict the onset of supercritical/subcritical transition within the hot-leg pipe.
Originalsprache | Englisch |
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Seiten | 5299-5314 |
Seitenumfang | 16 |
Publikationsstatus | Veröffentlicht - 2019 |
Veranstaltung | 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019 - Portland, USA/Vereinigte Staaten Dauer: 18 Aug. 2019 → 23 Aug. 2019 |
Konferenz
Konferenz | 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019 |
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Land/Gebiet | USA/Vereinigte Staaten |
Ort | Portland |
Zeitraum | 18/08/19 → 23/08/19 |