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Proposal of a novel CHF correlation for PWR under low pressure conditions based on stepwise regression method

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

Abstract

Constrained by limited experimental data, development of CHF correlation for PWR fuel assemblies under transient and accidental conditions at low pressure levels (2–10 MPa) is a typical statistical problem with small sample amounts, but simultaneously has requirements of high prediction accuracy. In this study, stepwise regression method was used to develop a new CHF correlation for application in PWR under low pressure conditions. First, several essential thermal-hydraulic parameters which might influence CHF were selected based on consensus characteristics of DNB phenomenon. With stepwise regression, the form and coefficients of the proposed CHF correlation were optimized in a dynamic manner. Compared to currently available CHF correlations, represented by the Westinghouse W-3 correlation, the CHF correlation obtained by stepwise regression has a much simpler form and matches also well with the experimental data.

Kurzfassung

Durch begrenzte experimentelle Daten gekennzeichniet stellt die Entwicklung einer CHF-Korrelation für DWR-Brennstoffelemente unter transienten und zufälligen Bedingungen bei niedrigen Drücken (2–10 MPa) ein typisches statistisches Problem bei kleinen Probenmengen dar. Gleichzeitig bestehen aber hohe Anforderungen an die Vorhersagegenauigkeit. In dieser Studie wurde mit Hilfe einer schrittweisen Regressionsmethode eine neue CHF-Korrelation für die Anwendung in DWR unter Niedrigdruckbedingungen entwickelt. Zunächst wurden einige wesentliche thermohydraulische Parameter, die den CHF beeinflussen können, auf der Grundlage von Konsenscharakteristika des DNB-Phänomens ausgewählt. Mit der schrittweisen Regression wurden Form und Koeffizienten der vorgeschlagenen CHF-Korrelation dynamisch optimiert. Im Vergleich zu den derzeit verfügbaren CHF-Korrelationen, dargestellt durch die Westinghouse W-3-Korrelation, hat die durch schrittweise Regression erhaltene CHF-Korrelation eine viel einfachere Form und passt dennoch gut zu den experimentellen Daten.

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References

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Received: 2018-01-02
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. Contents
  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. Improvements of interfacial friction and heat transfer models for rectangular narrow channel reflood simulation based on RELAP5
  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/10.3139/124.110881

Articles in the same Issue

  1. Contents/Inhalt
  2. Contents
  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. Improvements of interfacial friction and heat transfer models for rectangular narrow channel reflood simulation based on RELAP5
  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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