Analysis code development for the direct reactor auxiliary cooling system of the pool-type sodium-cooled fast reactor
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D. L. Zhang
Abstract
Sodium-cooled fast reactor (SFR) is one of the most promising reactors among the six Gen-IV reactor systems due to its significant advantages in close fuel cycle, comprehensive technology foundations and operation experiences. China is designing and constructing a demonstration SFR, in which, to ensure the reactor passive safety, a direct reactor auxiliary cooling system (DRACS) with the inter-wrapper flow is proposed as the decay heat removal system. Xi'an Jiaotong University is in charge of the DRACS analysis code development. The physical models in the DRACS are extracted and the numerical models for each component are established. The Gear method and the SIMPLE method are adopted as the primary solution algorithm. The code is developed and validated by EBR-II and PHENIX benchmarks, the results of which indicate that the code can predict experimental results very well.
Kurzfassung
Das Konzept des natriumgekühlten schnellen Reaktors SFR ist einer der vielversprechendsten Reaktoren unter den sechs Gen-IV-Reaktorsystemen aufgrund seiner signifikanten Vorteile im geschlossenen Brennstoffkreislauf, seiner umfassenden technologischen Grundlagen und Betriebserfahrungen. China plant und baut die Demonstrationsanlage SFR, in der, um die passive Sicherheit des Reaktors zu gewährleisten, ein direktes Reaktor-Hilfskühlsystem (DRACS) als Nachzerfallswärmeabfuhrsystem die Strömung zwischen den Abstandshaltern der Brennelemente vorgesehen wird. Die Xi'an Jiaotong Universität ist für die Entwicklung des DRACS-Analysecodes verantwortlich. Dazu werden die physikalischen Modelle im DRACS entwickelt und die numerischen Modelle für jede Komponente erstellt. Die Gear-Methode und die SIMPLE-Methode werden als primärer Lösungsalgorithmus übernommen. Der Code wird mit EBR-II und PHENIX-Benchmarks entwickelt und validiert, deren Ergebnisse zeigen, dass der Code experimentelle Ergebnisse sehr gut vorhersagen kann.
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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
