Simulation of water hammer phenomena using the system code ATHLET
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C. Bratfisch
Abstract
Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.
Kurzfassung
Druckstoßphänomene können sowohl in Kernkraftwerken als auch in verfahrenstechnischen Anlagen zu einer Gefährdung der Integrität von umliegenden Strukturen sowie zu einem möglichen Versagen von Rohrleitungen führen. In Kernkraftwerken kann dieses Phänomen im Verlauf einer Transiente oder eines hypothetischen Störfalls durch das Anfahren von Hilfs- sowie von Notkühlsystemen in Kombination mit schnell agierenden Armaturen auftreten. Zur weiteren Entwicklung und -validierung des Codes ATHLET (Analyse der Thermohydraulik von Lecks und Transienten) wird ein Experiment der Versuchsanlage Pilot Plant Pipework (PPP) des Fraunhofer UMSICHT mit der Version ATHLET 3.0A simuliert.
References
1 KSB Aktiengesellschaft: Der Druckstoß, Know-how Band 1; Information Brochure of the stock company KSB, Halle, 2013Search in Google Scholar
2 Dudlik, A.; Prasser, H.-M.: Water hammer and condensation hammer scenarios in power plants using new measurement system. Journal Forschung im Ingenieuerwesen73, Springer Verlag (2009), 10.1007/s10010-009-0100-9Search in Google Scholar
3 Neuhaus, T.: Mathematische Modellierung und vergleichende Untersuchungen zur Beschreibung von transienten Ein- und Mehrphasenströmungen in Rohrleitungen. Dissertation, Universität Dortmund, 2005Search in Google Scholar
© 2017, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- SIMULATE-3 K coupled code applications
- Application of the new IAPWS Guideline on the fast and accurate calculation of steam and water properties with the Spline-Based Table Look-Up Method (SBTL) in RELAP-7
- Simulation of water hammer phenomena using the system code ATHLET
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- Sensitivity analysis for thermo-hydraulics model of a Westinghouse type PWR: verification of the simulation results
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- The effect of boron dilution transient on the VVER-1000 reactor core using MCNP and COBRA-EN codes
- Modelling of the spent fuel heat-up in the spent fuel pools using one-dimensional system codes and CFD codes
- Optimization and analysis of the effects of physical parameters in a TRIGA-ADSR
- A comparison study for mass attenuation coefficients of some amino acids using MCNP code
- Validation of radioactive isotope activity measurement in homogeneous waste drum using Monte Carlo codes
- Study of the response reduction of LiF:Mg, Ti dosimeter for high dose dosimetry
- Non-contact micro mass evaluation method using an X-ray microscope
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- SIMULATE-3 K coupled code applications
- Application of the new IAPWS Guideline on the fast and accurate calculation of steam and water properties with the Spline-Based Table Look-Up Method (SBTL) in RELAP-7
- Simulation of water hammer phenomena using the system code ATHLET
- New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses
- Sensitivity analysis for thermo-hydraulics model of a Westinghouse type PWR: verification of the simulation results
- Calculation of the fuel temperature coefficient of reactivity considering non-uniform radial temperature distribution in the fuel rod
- The effect of boron dilution transient on the VVER-1000 reactor core using MCNP and COBRA-EN codes
- Modelling of the spent fuel heat-up in the spent fuel pools using one-dimensional system codes and CFD codes
- Optimization and analysis of the effects of physical parameters in a TRIGA-ADSR
- A comparison study for mass attenuation coefficients of some amino acids using MCNP code
- Validation of radioactive isotope activity measurement in homogeneous waste drum using Monte Carlo codes
- Study of the response reduction of LiF:Mg, Ti dosimeter for high dose dosimetry
- Non-contact micro mass evaluation method using an X-ray microscope
