Research on thermal-hydraulic behavior in the spent fuel pool using a full-height experimental facility
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Q. Cao
Abstract
During accident scenarios the effective cooling of spent fuel directly affects the safety of nuclear power plants. Two experiments were performed in a full-height facility to study the thermal-hydraulic behavior in spent fuel pool. In spent fuel pool boiling experiment, the heat transfer characteristics are related to the flow patterns. However, the flow pattern in narrow and long channel is different from the traditional flow pattern. In the semi-dry of heated rod, wall temperature oscillation occurs for a long time. In the spent fuel pray experiment, the liquid film thickness varies randomly with time and space. As the spray flow density increase, the maximum wall temperature decrease gradually with a certain linear characteristic.
Kurzfassung
Bei Unfallszenarien wirkt sich die effektive Kühlung abgebrannter Brennelemente direkt auf die Sicherheit von nuklearen Kraftwerken aus. Zur Untersuchung des thermohydraulischen Verhaltens im Brennelementlagerbecken wurden zwei Experimente in einer Versuchsanlage mit Stabbündeln in Originallänge durchgeführt. Beim Siedeversuch mit abgebrannten Brennelementen werden die Wärmeübertragungseigenschaften mit den Strömungsmustern in Beziehung gesetzt. Das Strömungsmuster in schmalen und langen Kanälen unterscheidet sich jedoch vom traditionellen Strömungsmuster. Im halbtrockenen Zustand des beheizten Stabes treten lange Zeit Wandtemperaturschwankungen auf. Im Experiment mit abgebrannten Brennelementen variiert die Dicke der Flüssigkeitsschicht zufällig mit der Zeit und dem Raum. Mit zunehmender Spritzflussdichte nimmt die maximale Wandtemperatur mit einer bestimmten linearen Kennlinie allmählich ab.
References
1 Chen, Y. S.; Yuan, Y. R.: Evaluation of cooling capacity with more fuel stored in the spent fuel pool of the Kuosheng plant Annals of Nuclear Energy109 (2017) 120–3410.1016/j.anucene.2017.05.026Suche in Google Scholar
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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
