Analysis of coolant flow in central tube of VVER-440 fuel assemblies
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G. Zsíros
Abstract
A CFD model has been developed to investigate the coolant flow in the central tube of VVER-440 fuel assemblies. The model has been validated with measured data of Kurchatov Institute. With this model a peripheral and an inner fuel assembly of the core have been investigated and the mass flux and heatup ratios of the central tube flow have been determined. Based on this study, the outlet mass flux of the tube is 1.2 times higher than the inlet mass flux of the rod bundle and heat-up in the tube is 0.35 times lower than the heat-up in the rod bundle. The ratios are not sensitive to the operational conditions within the scope of these investigations. The results of these simulations can be used as boundary conditions for the central tube in the assembly head calculations.
Kurzfassung
Zur Berechnung der Kühlmittelströmung im Zentralkanal eines WWER-440 Brennelementes wurde ein CFD Modell entwickelt. Die Validierung dieses Modells erfolgte mit Messdaten des Kurchatov Institutes. Im Beitrag werden Analysen der Kühlmittelströmung im Zentralkanals für ein innen und ein auβen im Reaktorkern angeordnetes Brennelement eines WWER-440 Kerns vorgestellt. Für diese wurden u. a. die Massenströme im Zentralkanal und die Aufheizung des Kühlmittels beim Durchströmen dieses Kanals berechnet. Daran schloss sich ein Vergleich der jeweiligen Parameter des Kanals mit den entsprechenden Parametern des Stabbündels an. Diese Analysen zeigen, dass der Austrittsmassenstrom des Kanals 1,2fach so groβ ist wie der Eintrittsmassenstrom des Stabbündels und die Aufheizung im Kanal 0,35fach kleiner ist als die Aufheizung des Stabbündels. Bei allen untersuchten Rechenläufen zeigte sich, dass diese Werte unabhängig von den jeweiligen Betriebsbedingungen waren. Die Ergebnisse dieser Berechnungen können als Randbedingungen des Zentralkanals bei Berechnungen der Strömungen im Brennelementkopf verwendet werden.
References
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© 2012, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries/Kurzfassungen
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – Selected contributions to the XXIst Symposium of the Atomic Energy Research organization
- Technical Contributions/Fachbeiträge
- Development of multi-group spectral code TVS-M
- Qualification of the APOLLO2 lattice physics code of the NURISP platform for VVER hexagonal lattices
- The simplified P3 approach on a trigonal geometry of the nodal reactor code DYN3D
- An analytical solution for the consideration of the effect of adjacent fuel assemblies; extension to VVER-440 type fuel assemblies
- Studies on boiling water reactor design with reduced moderation and analysis of reactivity accidents using the code DYN3D-MG
- Simulations of RUTA-70 reactor with CERMET fuel using DYN3D/ATHLET and DYN3D/RELAP5 coupled codes
- Analysis of coolant flow in central tube of VVER-440 fuel assemblies
- Effect of spacer grid mixing vanes on coolant outlet temperature distribution
- Study on severe accidents and countermeasures for VVER-1000 reactors using the integral code ASTEC
- Assessment of spectral history influence on PWR and WWER core
- New practice for the evaluation of rod efficiency measurement by rod drop at the NPP Paks
- Comparison of square and hexagonal fuel lattices for high conversion PWRs
- VVER-440 with inert matrix fuel – viable direction to sustainability
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries/Kurzfassungen
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – Selected contributions to the XXIst Symposium of the Atomic Energy Research organization
- Technical Contributions/Fachbeiträge
- Development of multi-group spectral code TVS-M
- Qualification of the APOLLO2 lattice physics code of the NURISP platform for VVER hexagonal lattices
- The simplified P3 approach on a trigonal geometry of the nodal reactor code DYN3D
- An analytical solution for the consideration of the effect of adjacent fuel assemblies; extension to VVER-440 type fuel assemblies
- Studies on boiling water reactor design with reduced moderation and analysis of reactivity accidents using the code DYN3D-MG
- Simulations of RUTA-70 reactor with CERMET fuel using DYN3D/ATHLET and DYN3D/RELAP5 coupled codes
- Analysis of coolant flow in central tube of VVER-440 fuel assemblies
- Effect of spacer grid mixing vanes on coolant outlet temperature distribution
- Study on severe accidents and countermeasures for VVER-1000 reactors using the integral code ASTEC
- Assessment of spectral history influence on PWR and WWER core
- New practice for the evaluation of rod efficiency measurement by rod drop at the NPP Paks
- Comparison of square and hexagonal fuel lattices for high conversion PWRs
- VVER-440 with inert matrix fuel – viable direction to sustainability
