Numerical and experimental investigation of 3D coolant temperature distribution in the hot legs of primary circuit of reactor plant with WWER-1000
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Yu. Saunin
Abstract
Nowadays the demands for accuracy of determining the weighted mean reactor power significantly increases due to implementing programs aimed at increasing installed power at nuclear power plants with WWER-1000 reactors. This accuracy strongly depends on the coolant temperature stratification in the primary circuit pipes, especially in the hot legs. Thus, the acute problem is to investigate the phenomenon on the basis of the full-scale tests experimental data and the calculated data of computational fluid dynamics simulations. The paper presents the full-scale experimental data on the coolant temperature distribution in the hot legs obtained via I&C systems, primarily via in-core monitoring system. These data were obtained at the pilot operation stage of Unit 4 Kalinin nuclear power plant in a stationary mode at nominal power level. To present the results of calculating simulation for the same mode CFD codes are used.
Kurzfassung
Infolge von Leistungserhöhungen in WWER-1000-Reaktoren steigen die Anforderungen an die Genauigkeiten bei der Berechnung der gewichteten mittleren Reaktorleistung signifikant. Diese Genauigkeit hängt stark von der Temperaturschichtung des Kühlmittels im Primärkreislauf, insbesondere im heißen Strang, ab. So stellt sich für die analytischen Untersuchungen die Aufgabenstellung die Phänomene auf der Basis von Full-Scale-Experimentdaten und CFD-Berechnungen zu untersuchen. In diesem Beitrag werden experimentelle Full-Scale-Daten der Temperaturverteilung des Kühlmittels im heißen Strand vorgestellt, die mit Hilfe von In-Kern-Monitoring-Systemen während des Pilotbetriebs des Kalinin 4 NPP bei stationären Betriebsbedingungen bei Nennleistung aufgenommen wurden. Diese Daten werden mit Ergebnissen aus CFD-Analysen gegenübergestellt.
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© 2015, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2014
- Technical Contributions/Fachbeiträge
- Assessment of the uncertainties of MULTICELL calculations by the OECD NEA UAM PWR pin cell burnup benchmark
- Development of codes and KASKAD complex
- Applying full multigroup cell characteristics from MCU code to finite difference calculations of neutron field in VVER core
- Calculations of 3D full-scale VVER fuel assembly and core models using MCU and BIPR-7A codes
- An analysis of reactivity prediction during the reactor start-up process
- Experimental and computational investigations of heat and mass transfer of intensifier grids
- Implementation of CFD module in the KORSAR thermal-hydraulic system code
- Numerical and experimental investigation of 3D coolant temperature distribution in the hot legs of primary circuit of reactor plant with WWER-1000
- Analyses of Beyond Design Basis Accident Homogeneous Boron Dilution Scenarios
- Analysis of heterogeneous boron dilution transients during outages with APROS 3D nodal core model
- Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle
- Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations
- Neutron-kinetic and thermo-hydraulic uncertainties in the study of Kalinin-3 benchmark
- Inter-assembly gap deviations in VVER-1000: Accounting for effects on engineering margin factors
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2014
- Technical Contributions/Fachbeiträge
- Assessment of the uncertainties of MULTICELL calculations by the OECD NEA UAM PWR pin cell burnup benchmark
- Development of codes and KASKAD complex
- Applying full multigroup cell characteristics from MCU code to finite difference calculations of neutron field in VVER core
- Calculations of 3D full-scale VVER fuel assembly and core models using MCU and BIPR-7A codes
- An analysis of reactivity prediction during the reactor start-up process
- Experimental and computational investigations of heat and mass transfer of intensifier grids
- Implementation of CFD module in the KORSAR thermal-hydraulic system code
- Numerical and experimental investigation of 3D coolant temperature distribution in the hot legs of primary circuit of reactor plant with WWER-1000
- Analyses of Beyond Design Basis Accident Homogeneous Boron Dilution Scenarios
- Analysis of heterogeneous boron dilution transients during outages with APROS 3D nodal core model
- Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle
- Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations
- Neutron-kinetic and thermo-hydraulic uncertainties in the study of Kalinin-3 benchmark
- Inter-assembly gap deviations in VVER-1000: Accounting for effects on engineering margin factors
