Investigation of condensation process at COSMEA test facility with ATHLET code
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Y. Zhang
Abstract
Safety is an essential topic in the development process of nuclear power plant. Several Generation III and III+ reactor designs contain passive safety system to control accident without external power. An example is the Emergency Condenser (EC) of the KERENA reactor design. The EC removes heat from the Reactor Pressure Vessel in the case of design accidents. The experimental facility COSMEA at Helmhotz Zentrum Dresden Rossendorf (HZDR) was set up to investigate the flow morphology and heat transfer structure of condensation inside a slightly inclined tube. In this paper, the condensation process in COSMEA was simulated with the thermal hydraulic system codes ATHLET. The performance of the ATHLET heat transfer models were identified. The simulation results were compared against the experiments. The heat flux, condensation rate and temperature of cooling water during the condensation was analyzed.
Kurzfassung
Sicherheit ist ein wesentliches Thema im Entwicklungsprozess von Kernkraftwerken. Mehrere Reaktortypen der Generation III und III+ enthalten ein passives Sicherheitssystem zur Unfallkontrolle ohne externe Stromversorgung. Ein Beispiel ist der Notkondensator (EC) des KERENA-Reaktorkonzeptes. Der EC entzieht dem Reaktordruckbehälter bei Störfällen Wärme. Die Versuchsanlage COSMEA am Helmholtz Zentrum Dresden Rossendorf (HZDR) wurde eingerichtet, um die Strömungsmorphologie und die Wärmeübertragungsstruktur der Kondensation in einem leicht geneigten Rohr zu untersuchen. In diesem Beitrag werden Nachrechnung des Kondensationsprozesses in der Versuchsanlage COSMEA mit den thermohydraulischen Systemcode ATHLET vorgestellt. Die Leistungsfähigkeit der ATHLET Wärmeübertragungsmodelle wurde bestimmt. Die Simulationsergebnisse wurden mit den Experimenten verglichen. Der Wärmefluss, die Kondensationsrate und die Temperatur des Kühlwassers während der Kondensation wurden analysiert.
References
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© 2018, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Editorial
- 2nd Sino-German Symposium on Fundamentals of Advanced Nuclear Safety Technology
- Technical Contributions/Fachbeiträge
- Challenging issues and recent R&D activities for enhancing nuclear safety in Korea
- Necessary improvements of the GRS simulation chain for the simulation of light-water-cooled SMRs
- Scaling analysis of core pressure drop in reduced height integral test facility
- Criteria and comparison of thermal stratification between PRHR HX heating and ADS spraying process in IRWST based on a down-scaled experimental facility
- Numerical simulation of bubble growth on and departure from the heated surface by an improved lattice Boltzmann model
- Improvements of interfacial friction and heat transfer models for rectangular narrow channel reflood simulation based on RELAP5
- Proposal of a novel CHF correlation for PWR under low pressure conditions based on stepwise regression method
- Mechanistic prediction of post dryout heat transfer and rewetting
- Investigation of condensation process at COSMEA test facility with ATHLET code
- Research on thermal-hydraulic behavior in the spent fuel pool using a full-height experimental facility
- Experimental investigation on the distribution of spray water in a spent fuel-assembly simulator
- Wet resuspension modelling and validation
- Analysis code development for the direct reactor auxiliary cooling system of the pool-type sodium-cooled fast reactor
- Comparison of heat transfer models with databank of supercritical fluid
- Blankets – key element of a fusion reactor – functions, design and present state of development
- Preliminary steady and transient analysis for the CFETR helium cooled solid blanket system with RELAP
- A methodology for thermo-mechanical assessment of in-box LOCA events on fusion blankets and its application to EU DEMO HCPB breeding blanket
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Editorial
- 2nd Sino-German Symposium on Fundamentals of Advanced Nuclear Safety Technology
- Technical Contributions/Fachbeiträge
- Challenging issues and recent R&D activities for enhancing nuclear safety in Korea
- Necessary improvements of the GRS simulation chain for the simulation of light-water-cooled SMRs
- Scaling analysis of core pressure drop in reduced height integral test facility
- Criteria and comparison of thermal stratification between PRHR HX heating and ADS spraying process in IRWST based on a down-scaled experimental facility
- Numerical simulation of bubble growth on and departure from the heated surface by an improved lattice Boltzmann model
- Improvements of interfacial friction and heat transfer models for rectangular narrow channel reflood simulation based on RELAP5
- Proposal of a novel CHF correlation for PWR under low pressure conditions based on stepwise regression method
- Mechanistic prediction of post dryout heat transfer and rewetting
- Investigation of condensation process at COSMEA test facility with ATHLET code
- Research on thermal-hydraulic behavior in the spent fuel pool using a full-height experimental facility
- Experimental investigation on the distribution of spray water in a spent fuel-assembly simulator
- Wet resuspension modelling and validation
- Analysis code development for the direct reactor auxiliary cooling system of the pool-type sodium-cooled fast reactor
- Comparison of heat transfer models with databank of supercritical fluid
- Blankets – key element of a fusion reactor – functions, design and present state of development
- Preliminary steady and transient analysis for the CFETR helium cooled solid blanket system with RELAP
- A methodology for thermo-mechanical assessment of in-box LOCA events on fusion blankets and its application to EU DEMO HCPB breeding blanket
