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Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled Molten Salt Reactors

  • S. Qiu , D. Zhang , L. Liu , M. Liu , R. Xu , C. Gong and G. H. Su
Published/Copyright: April 19, 2016
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Abstract

Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

Kurzfassung

Wegen ihres Brennstoffkreislaufs und der Thorium-Nutzung sind Salzschmelze-Reaktoren in China von besonderem Interesse. Die Nutzung des flüssigen Brennstoffs führt zu komplizierteren gekoppelten Berechnungen der Neutronenkinetik und der Thermohydraulik im Vergleich zu Feststoffbrennstoff. In diesem Beitrag werden die Grundlagen zur Berechnung und Analyse der Sicherheit von Salzschmelze-Reaktoren mit flüssigem Brennstoff beschrieben. Diese basieren auf dem MOSART Konzept. Die im stationären Zustand bestimmten Kenndaten bilden dann die Grundlage für Sicherheitsanalysen. Am Beispiel der Berechnung einer Loss of Flow Transiente werden die inhärenten Sicherheitseigenschaften des MOSART Konzepts aufgrund seines streng negativen Reaktivitäts-Feedbacks dargestellt.


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Received: 2015-12-16
Published Online: 2016-04-19
Published in Print: 2016-04-27

© 2016, Carl Hanser Verlag, München

Articles in the same Issue

  1. Contents/Inhalt
  2. Contents
  3. Summaries/Kurzfassungen
  4. Summaries
  5. Editorial
  6. Selected contributions from 1th Sino-German Symposium on Fundamentals of Advanced Nuclear Safety Technology
  7. Technical Contributions/Fachbeiträge
  8. Scientific codes developed and used at GRS – Nuclear simulation chain
  9. Challenges on innovations of newly-developed safety analysis codes
  10. Validation of system codes for plant application on selected experiments
  11. Progress of Experimental Research on Nuclear Safety in NPIC
  12. Severe accident research activities at Helmholtz-Zentrum Dresden-Rossendorf (HZDR)
  13. THAI experimental programme for containment safety assessment under severe accident conditions
  14. A spray cooling technique for spent fuel assembly stored in pool
  15. KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors
  16. Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled Molten Salt Reactors
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  18. Validation of the ATHLET-SC code by trans-critical transient data
  19. Qualification of CFD-models for multiphase flows
  20. The reactor dynamics code DYN3D
  21. Critical flow phenomena and modeling in advanced nuclear safety technology
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