Validation of the ATHLET-SC code by trans-critical transient data
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X. Liu
Abstract
For the safety analysis of Supercritical Water-Cooled Reactor (SCWR), one of the challenge tasks is to predict the trans-critical behavior of the reactor system during some accidents. The current safety codes have some shortcomings when the pressure decreases from the supercritical condition to the subcritical state due to the void fraction discontinuity across the critical point. Another challenge is the validation of the system code, which needs the transient experimental data. To overcome the above-mentioned challenges, this paper validates the modified code ATHLET-SC, which is developed based on the pseudo two-phase method. The trans-critical transient data from SWAMUP test facility in Shanghai Jiao Tong University (SJTU) are adopted to compare with the simulation results. The results obtained so far shows that the ATHLET-SC code has good feasibility to the trans-critical simulation of SCWR, and it can be used for transient analysis of SCWR in the future.
Kurzfassung
In Sicherheitsanalysen zum Supercritical Water-Cooled Reactor (SCWR) eine Hauptaufgabe ist die Vorhersage des trans-kritischen Verhaltens des Reaktorsystems während des Störfallablaufs. Aktuelle Systemcodes berücksichtigen den Druckabfall von überkritischen zu unterkritischen Zuständen infolge der Unstetigkeit des Dampfgehalts oberhalb des kritischen Punktes nicht ausreichend. Auch standen zur Validierung bislang nicht ausreichend Experimentdaten zur Verfügung. In diesem Beitrag wird eine Erweiterung für den Systemcode ATHLET basierend auf der pseudo Zweiphasenmethode vorgestellt. Rechnungen mit dieser ATHLET-SC genannten Version wurden mit experimentellen Daten der Versuchsanlage SWAMUP der Shanghai Jiao Tong University (SJTU) verglichen. Basierend auf den dabei erzielten Ergebnissen soll diese Erweiterung auch bei Transientenanalysen mit trans-kritischen Zuständen im SCWR verwendet werden.
References
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2 Fu, S. W.; Liu, X. J.; Zhou, C.; et al.: Modification and application of the system analysis code ATHLET to trans-critical simulations. Annals of Nuclear Energy44 (2012) 4010.1016/j.anucene.2012.02.005Search in Google Scholar
3 Zhao, M.; Gu, H. Y.; Li, H. B.; et al.: Experimental and numerical study on heat transfer of supercritical water flowing upward in 2 × 2 rod bundles. Proc. Int. Conf. the 16th International Topical Meeting on Nuclear Reactor Thermalhydraulics (NURETH-16), Chicago, USA, August 30-September 04, 2015Search in Google Scholar
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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
- 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
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- 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
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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
