Critical flow phenomena and modeling in advanced nuclear safety technology
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Y. Z. Chen
Abstract
The discharge could be non-choking or choking, depending on the break shape, length and conditions. This presents a challenge in the calculation of standard problems. A stable experiment of water was performed to study the break flow rate in nozzles of diameter of 1.41 and 2.0 mm with rounded-edge and sharp-edge. The pressure covered the ranges of 0.5 to 29.5 MPa, inlet quality 0 to 1.0 and subcooling up to 350 °C. The results exhibited a close relation of thermal non-equilibrium with pressure. For supercritical pressure a modified equilibrium model in combination with the Bernoulli equation is presented.
Kurzfassung
Das Verhalten von Flüssigkeiten beim Ausströmen hängt direkt von der Geometrie der Öffnung sowie von den dort herrschenden Betriebsbedingungen ab. Zur Entwicklung und Überprüfung von Berechnungsmodellen wurde ein Experiment mit einer Wasserausströmung aus einem Bruchquerschnitt durchgeführt. Dabei wurde der Massenstrom aus gerundeten und scharfkantigen Düsendurchmessern von 1.41 und 2.0 mm untersucht. Der Druckbereich lag zwischen 0.5 und 29.5 MPa, die Unterkühlung bei bis zu 350 °C und der Eintrittsgasgehalt zwischen 0 und 1.0. Die Ergebnisse zeigten einen starken Zusammenhang zwischen thermischem Nichtgleichgewicht und Druck. Für überkritische Drücke wird ein modifiziertes Gleichgewichtsmodell unter Berücksichtigung der Bernoulli-Gleichung präsentiert.
References
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© 2016, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Selected contributions from 1th Sino-German Symposium on Fundamentals of Advanced Nuclear Safety Technology
- Technical Contributions/Fachbeiträge
- Scientific codes developed and used at GRS – Nuclear simulation chain
- Challenges on innovations of newly-developed safety analysis codes
- Validation of system codes for plant application on selected experiments
- Progress of Experimental Research on Nuclear Safety in NPIC
- Severe accident research activities at Helmholtz-Zentrum Dresden-Rossendorf (HZDR)
- THAI experimental programme for containment safety assessment under severe accident conditions
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- Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled Molten Salt Reactors
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- Validation of the ATHLET-SC code by trans-critical transient data
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Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Selected contributions from 1th Sino-German Symposium on Fundamentals of Advanced Nuclear Safety Technology
- Technical Contributions/Fachbeiträge
- Scientific codes developed and used at GRS – Nuclear simulation chain
- Challenges on innovations of newly-developed safety analysis codes
- Validation of system codes for plant application on selected experiments
- Progress of Experimental Research on Nuclear Safety in NPIC
- Severe accident research activities at Helmholtz-Zentrum Dresden-Rossendorf (HZDR)
- THAI experimental programme for containment safety assessment under severe accident conditions
- A spray cooling technique for spent fuel assembly stored in pool
- KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors
- Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled Molten Salt Reactors
- 10.3139/124.110680
- Validation of the ATHLET-SC code by trans-critical transient data
- Qualification of CFD-models for multiphase flows
- The reactor dynamics code DYN3D
- Critical flow phenomena and modeling in advanced nuclear safety technology
- 10.3139/124.110682
- Safety and security aspects in design of digital safety I&C in nuclear power plants
- Thermohydraulic safety issues for liquid metal cooled systems
- Design and safety analysis of the helium cooled solid breeder blanket for CFETR
- Qualification of pebble fuel for HTGRs
- High temperature reactors for cogeneration applications
