Methodology for neutronic uncertainty propagation and application to a UAM-LWR benchmark
In: Progress in nuclear energy: the international review journal covering all aspects of nuclear energy, Band 126, S. 103389
ISSN: 0149-1970
17 Ergebnisse
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In: Progress in nuclear energy: the international review journal covering all aspects of nuclear energy, Band 126, S. 103389
ISSN: 0149-1970
In: Progress in nuclear energy: the international review journal covering all aspects of nuclear energy, Band 148, S. 104233
ISSN: 0149-1970
In: Progress in nuclear energy: the international review journal covering all aspects of nuclear energy, Band 127, S. 103460
ISSN: 0149-1970
In: Progress in nuclear energy: the international review journal covering all aspects of nuclear energy, Band 115, S. 181-193
ISSN: 0149-1970
In: Progress in nuclear energy: the international review journal covering all aspects of nuclear energy, Band 53, Heft 8, S. 1136-1139
ISSN: 0149-1970
In: Progress in nuclear energy: the international review journal covering all aspects of nuclear energy, Band 43, Heft 1-4, S. 187-194
ISSN: 0149-1970
In: Progress in nuclear energy: the international review journal covering all aspects of nuclear energy, Band 135, S. 103701
ISSN: 0149-1970
In: Progress in nuclear energy: the international review journal covering all aspects of nuclear energy, Band 127, S. 103463
ISSN: 0149-1970
In: Progress in nuclear energy: the international review journal covering all aspects of nuclear energy, Band 43, Heft 1-4, S. 329-336
ISSN: 0149-1970
In: Progress in nuclear energy: the international review journal covering all aspects of nuclear energy, Band 36, Heft 4, S. 469
ISSN: 0149-1970
In: Progress in nuclear energy: the international review journal covering all aspects of nuclear energy, Band 36, Heft 2, S. 189-229
ISSN: 0149-1970
In: Progress in nuclear energy: the international review journal covering all aspects of nuclear energy, Band 29, Heft 3-4, S. 321-336
ISSN: 0149-1970
In: Progress in nuclear energy: the international review journal covering all aspects of nuclear energy, Band 115, S. 169-180
ISSN: 0149-1970
In: Progress in nuclear energy: the international review journal covering all aspects of nuclear energy, Band 53, Heft 8, S. 1084-1090
ISSN: 0149-1970
In: Science and technology of nuclear installations, Band 2018, S. 1-10
ISSN: 1687-6083
The one-dimensional two-fluid model approach has been traditionally used in thermal-hydraulics codes for the analysis of transients and accidents in water–cooled nuclear power plants. This paper investigates the performance of RELAP5/MOD3 predicting vertical upward bubbly flow at low velocity conditions. For bubbly flow and vertical pipes, this code applies the drift-velocity approach, showing important discrepancies with the experiments compared. Then, we use a classical formulation of the drag coefficient approach to evaluate the performance of both approaches. This is based on the critical Weber criteria and includes several assumptions for the calculation of the interfacial area and bubble size that are evaluated in this work. A more accurate drag coefficient approach is proposed and implemented in RELAP5/MOD3. Instead of using the Weber criteria, the bubble size distribution is directly considered. This allows the calculation of the interfacial area directly from the definition of Sauter mean diameter of a distribution. The results show that only the proposed approach was able to predict all the flow characteristics, in particular the bubble size and interfacial area concentration. Finally, the computational results are analyzed and validated with cross-section area average measurements of void fraction, dispersed phase velocity, bubble size, and interfacial area concentration.