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Scaling analysis of core pressure drop in reduced height integral test facility

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Published/Copyright: June 8, 2018
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Kerntechnik
From the journal Kerntechnik Volume 83 Issue 3

Abstract

Integral test plays essential role to assess the design of the emergency cooling system of nuclear reactors. Different from full height integral test facilities, reduced height integral test facilities have new problems on the pressure drop scaling. This paper mainly focuses on scaling of pressure drop across the core as it is the major pressure drop in primary loop. The analysis of pressure drop across the core has been divided into three terms and each term has been discussed separately based on two conditions: the normal operation condition and natural circulation condition. After that, the total pressure drop ratios under these two conditions have been discussed.

Kurzfassung

Integraltests spielen eine wesentliche Rolle bei der Beurteilung der Auslegung des Notkühlsystems von Kernkraftwerken. Anders als bei nicht höhenskalierten Integralversuchsanlagen führt die Reduktion der Höhe in Integralversuchsanlagen zu neuen Problemen bei der Skalierung des Druckverlustes. Dieses Papier konzentriert sich auf die Skalierung des Druckabfalls über dem Kern, da dieser den größten Druckabfall im Primärkreislauf liefert. Die Analyse des Druckabfalls über dem Kern wird in drei Teile unterteilt und jeder Teil wird separat unter Berücksichtigung des normalen Betriebszustand und des natürlichen Kreislaufzustands diskutiert. Daran anschließend werden die Gesamtdruckverluste bestimmt und diskutiert.

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References

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Received: 2018-01-16
Published Online: 2018-06-08
Published in Print: 2018-06-18

© 2018, Carl Hanser Verlag, München

Articles in the same Issue

  1. Contents/Inhalt
  2. 10.3139/124.018032
  3. Editorial
  4. 2nd Sino-German Symposium on Fundamentals of Advanced Nuclear Safety Technology
  5. Technical Contributions/Fachbeiträge
  6. Challenging issues and recent R&D activities for enhancing nuclear safety in Korea
  7. Necessary improvements of the GRS simulation chain for the simulation of light-water-cooled SMRs
  8. Scaling analysis of core pressure drop in reduced height integral test facility
  9. Criteria and comparison of thermal stratification between PRHR HX heating and ADS spraying process in IRWST based on a down-scaled experimental facility
  10. Numerical simulation of bubble growth on and departure from the heated surface by an improved lattice Boltzmann model
  11. 10.3139/124.110898
  12. Proposal of a novel CHF correlation for PWR under low pressure conditions based on stepwise regression method
  13. Mechanistic prediction of post dryout heat transfer and rewetting
  14. Investigation of condensation process at COSMEA test facility with ATHLET code
  15. Research on thermal-hydraulic behavior in the spent fuel pool using a full-height experimental facility
  16. Experimental investigation on the distribution of spray water in a spent fuel-assembly simulator
  17. Wet resuspension modelling and validation
  18. Analysis code development for the direct reactor auxiliary cooling system of the pool-type sodium-cooled fast reactor
  19. Comparison of heat transfer models with databank of supercritical fluid
  20. Blankets – key element of a fusion reactor – functions, design and present state of development
  21. Preliminary steady and transient analysis for the CFETR helium cooled solid blanket system with RELAP
  22. A methodology for thermo-mechanical assessment of in-box LOCA events on fusion blankets and its application to EU DEMO HCPB breeding blanket
https://doi.org/10.3139/124.110897

Articles in the same Issue

  1. Contents/Inhalt
  2. 10.3139/124.018032
  3. Editorial
  4. 2nd Sino-German Symposium on Fundamentals of Advanced Nuclear Safety Technology
  5. Technical Contributions/Fachbeiträge
  6. Challenging issues and recent R&D activities for enhancing nuclear safety in Korea
  7. Necessary improvements of the GRS simulation chain for the simulation of light-water-cooled SMRs
  8. Scaling analysis of core pressure drop in reduced height integral test facility
  9. Criteria and comparison of thermal stratification between PRHR HX heating and ADS spraying process in IRWST based on a down-scaled experimental facility
  10. Numerical simulation of bubble growth on and departure from the heated surface by an improved lattice Boltzmann model
  11. 10.3139/124.110898
  12. Proposal of a novel CHF correlation for PWR under low pressure conditions based on stepwise regression method
  13. Mechanistic prediction of post dryout heat transfer and rewetting
  14. Investigation of condensation process at COSMEA test facility with ATHLET code
  15. Research on thermal-hydraulic behavior in the spent fuel pool using a full-height experimental facility
  16. Experimental investigation on the distribution of spray water in a spent fuel-assembly simulator
  17. Wet resuspension modelling and validation
  18. Analysis code development for the direct reactor auxiliary cooling system of the pool-type sodium-cooled fast reactor
  19. Comparison of heat transfer models with databank of supercritical fluid
  20. Blankets – key element of a fusion reactor – functions, design and present state of development
  21. Preliminary steady and transient analysis for the CFETR helium cooled solid blanket system with RELAP
  22. A methodology for thermo-mechanical assessment of in-box LOCA events on fusion blankets and its application to EU DEMO HCPB breeding blanket
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