Title of article
Simulation of subcooled flow instability for high flux research reactors using the extended code ATHLET Original Research Article
Author/Authors
A Hainoun، نويسنده , , A Schaffrath، نويسنده ,
Issue Information
روزنامه با شماره پیاپی سال 2001
Pages
18
From page
163
To page
180
Abstract
Covering the wide range of reactor safety analysis of power reactors, consisting of leak and transients, the thermohydraulic code ATHLET is being developed by the Gesellschaft for Anlagen- und Reaktorsicherheit (GRS-Society for Plant and Reactor Safety) (Lerchel, G., Austregesilo, H., 1998. ATHLET Mode 1.2 Cycle A, Userʹs Manual, GRS-p-1/Vol. 1, Rev. 1, GRS). In order to extend the codeʹs range of application to the safety analysis of research reactors, a model was developed and implemented permitting a description of the steam formation in the subcooled boiling regime (Hainoun, A., 1994. Modellierung des unterkühlten Siedens in ATHLET und Anwendung in wassergekühlten Forschungsreaktoren, D 294 Diss. Univ. Bochum, Jül-2961). Considering the specific features of high flux research reactors given by both high heat flux and high flow velocity, the model of void condensation in subcooled flow has been extended and a new correlation of critical heat flux (CHF) is implemented. To validate the extended program, the Thermal–Hydraulic Test Loop (THTL) of Oak Ridge National Laboratory (ORNL) was modeled with ATHLET and an extensive series of experiments concerning the onset of thermohydraulic flow instability (OFI) in subcooled boiling regime were calculated. The comparison between experiments and ATHLET-postcalculation shows that the extended code can accurately simulate the thermohydraulic conditions of flow instability in a wide range of heat flux up to 15 MW m−2 and inlet flow velocity up to 20 m s−1. The thermohydraulic design limit characterized by the mass flux, at which the flow just becomes unstable (OFI), has been predicted in very good agreement with the experiment. However the calculated pressure drop at OFI is overestimated by a maximum deviation of about 25%. The calculated exit bulk temperature of subcooled coolant and the maximum wall temperature at OFI show a maximum deviation from experiment of 12 and 7%, respectively.
Journal title
Nuclear Engineering and Design Eslah
Serial Year
2001
Journal title
Nuclear Engineering and Design Eslah
Record number
889364
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