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.
References
1 Sokolov, D. A.; Kim, V. V.; Kuznetcov, V. I.: Increase of power WWER-1000. Works of the Odessa polytechnical university, v. 2 (28), 2007, (in Russian)Suche in Google Scholar
2 Asmolov, V. G.: Maintenance of safety and increase of efficiency of atomic engineering of Russian. Proc. 9th Int. Conf. Safety, efficiency and atomic engineering economy, JSC “Concern Rosenergoatom”, Moscow, May 21–23, 2014, (in Russian)Suche in Google Scholar
3 Saunin, Yu. V.; Dobrotvorsky, A. N.; Semenikhin, A. V.: Examination of factors determining the coolant thermal stratification in hot legs of primary circuit loops of NPPs with WWER-1000 reactors. Proc. 8th Int. Conf. Safety, Assur. Nucl. Power Plants with WWER, JSC EDO “Gidropress”, Podolsk, 2013 (in Russian)Suche in Google Scholar
4 Bai, V. F.; Bogachek, L. N; Makarov, S. V.; Lupishko, A. N.: The state of the in-core temperature monitoring and the analysis of basic thermophysical power plant characteristics at Kalinin NPP power units. Proc. 7th Int. Conf. Safety, efficiency and atomic engineering economy, JSC “Concern Rosenergoatom”, Moscow, 2010 (in Russian)Suche in Google Scholar
5 Hashemian, H. M.: Maintenance of process instrumentation in nuclear power plants. Springer-VerlagBerlin Heidelberg, 2006Suche in Google Scholar
6 Chiang, J. S. C.; Pei, B. S.; Tsai, F. P.: Pressurized water reactor (PWR) hot-leg streaming. Part 1: computational fluid dynamics (CFD) simulations. Nuclear Engineering and Design241 (2011) 1768–177510.1016/j.nucengdes.2009.12.028Suche in Google Scholar
7 Smith, B. L.: Assessment of CFD codes used in nuclear reactor safety simulations. Nuclear Engineering and Technology42 (2010) 10.5516/net.2010.42.4.339Suche in Google Scholar
8 Kulish, G. V.; Abdullaev, A. M. etal.: 3D distribution of the coolant temperature in PCP of the power unit No2 South Ukrainian NPP. Proc. 8th Int. Conf. Safety, Assur. Nucl. Power Plants with WWER. JSC EDO “Gidropress”, Podolsk, 2013 (in Russian)Suche in Google Scholar
9 Kulish, G. V.: Calculation of the coolant flow distribution in a core of the power unit No3 South Ukrainian NPP in 21st-24th fuel lifetimes. CN-LFD-08-07, Edition 0, CDA, 2007 (in Russian)Suche in Google Scholar
10 Saunin, Yu. V.; Dobrotvorsky, A. N.; Semenikhin, A. V.: Research of the coolant temperature at a core inlet at commissioning the power unit No. 2 Rostov NPP. Proc. 7th Int. Conf. Safety, Assur. Nucl. Power Plants with WWER. JSC EDO “Gidropress”, Podolsk, 2011 (in Russian)Suche in Google Scholar
11 SolidWorks® Office Premium 2009×64 Edition. Reference GuideSuche in Google Scholar
12 Bottcher, M.: Detailed CFX-5 study of the coolant mixing within the reactor pressure vessel of a VVER-1000 reactor during a non-symmetrical heat-up test. Nuclear Engineering and Design238 (2008) 445–45210.1016/j.nucengdes.2007.02.054Suche in Google Scholar
13 Prasser, H.-M.; Kliem, S.: Coolant mixing experiments in the upper plenum of the ROCOM test facility. Nuclear Engineering and Design276 (2014) 30–4210.1016/j.nucengdes.2014.05.016Suche in Google Scholar
14 Bieder, U. etal.: Simulation of mixing effects in a VVER-1000 reactor. Nuclear Engineering and Design237 (2007) 171810.1016/j.nucengdes.2007.02.015Suche in Google Scholar
15 Lisenkov, E. etal.: Investigation of coolant mixing in reactor WWER-1000. Proc. 6th Int. Conf. Safety, Assur. Nucl. Power Plants with WWER. JSC EDO “Gidropress”, Podolsk, 2009 (in Russian)Suche in Google Scholar
© 2015, Carl Hanser Verlag, München
Artikel in diesem Heft
- 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
Artikel in diesem Heft
- 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
