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Heat transfer study of a submerged reactor channel under boil-off condition

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Published/Copyright: May 18, 2013
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Kerntechnik
From the journal Kerntechnik Volume 77 Issue 6

Abstract

Experiments have been carried out to study the heatup behavior of a single segmented reactor channel for Pressurized Heavy Water Reactor under submerged, partially submerged and exposed conditions. This situation may arise from a severe accident scenario of Pressurised Heavy Water Reactors where full or segmented reactor channels are likely to be disassembled and form a submerged debris bed. An assembly of electrical heater rod, simulating fuel bundle and channel components like Pressure Tube and Calandria Tube constitutes the segmented reactor channel. Heatup of this assembly is observed with respect to different water levels ranging from full submergence to totally exposed and power levels of 6–8 kW, typical to decay power level. It has been observed from the set of experiment that fuel bundle local dry out followed by heatup does not happen till the bundle is partially submerged. Temperature excursion of the bundle is evident when the bundle is exposed to steam-air environment.

Kurzfassung

Für einen einzelnen Reaktorkanal eines schwerwassermoderierten Druckreaktor (PHWR) wurden Experimente zur Untersuchung des Aufheizverhaltens dieses Kanals unter den drei während eines schweren Störfallszenariums möglichen Bedingungen vollständig überflutet, teilweise überflutet und freiliegend. Im Experiment wird das Brennelementsegment durch elektrische Heizstäbe zur Simulation des Brennelements und der Komponenten Druckrohr und Calandriarohr nachgebildet. Das Aufheizverhalten wird bei verschiedenen Wasserständen von vollkommen freiliegend bis vollkommen überflutet untersucht. Dabei werden typische Nachzerfallsleistungen zwischen 6 und 8 kW aufgebracht. So wurde experimentell festgestellt, dass das lokale Austrocknen des Brennelements infolge der Aufheizung nicht auftritt, solange das Bündel zumindest teilweise überflutet ist. Sobald das Bündel einer Wasserdampf-Luft-Umgebung ausgesetzt ist, findet erwartungsgemäß eine Temperaturanstieg statt.

References

1 Analysis of Severe Accidents in Pressurized Heavy Water Reactors, IAEA TECDOC1594 (2008)Search in Google Scholar

2 Mathew, P. M.; Kupferschmidt, W. C. H.; Snell, V. G.; Bonechi, M.: CANDU-Specific Severe Core Damage Accident Experiments in Support of Level 2 PSA. Proc. 16th Int. Conf.on Structural Mechanics in Reactor Technology, SMiRT 16, (2001)Search in Google Scholar

3 Mathew, P. M.; White, A. J.; Snell, V. G.; Bonechi, M.: Severe Core Damage Accident Analyses and Experiments for CANDU Applications. Proc. 17th Int. Conf. on Structural Mechanics in Reactor Technology, SMiRT17 (2003)Search in Google Scholar

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Received: 2012-7-20
Published Online: 2013-05-18
Published in Print: 2012-12-01

© 2012, Carl Hanser Verlag, München

Articles in the same Issue

  1. Contents/Inhalt
  2. Contents
  3. Summaries/Kurzfassungen
  4. Summaries
  5. Technical Contributions/Fachbeiträge
  6. Remarks on boiling water reactor stability analysis – part 2: stability monitoring
  7. Irradiation tests on PHWR type fuel elements in TRIGA research reactor of INR Pitesti
  8. Heat transfer study of a submerged reactor channel under boil-off condition
  9. Dynamic assessment for life extension of nuclear power plants (NPPs) using system dynamics (SD) method
  10. Effect of using FLiBe and FLiNaBe molten salts bearing plutonium fluorides on the neutronic performance of PACER
  11. Evaluations of the CCFL and critical flow models in TRACE for PWR LBLOCA analysis
  12. Development of program DETSIM to simulate detector's full energy peak efficiency
  13. The criticality calculations for one-speed neutrons in a reflected slab with anisotropic scattering using the modified UN method
  14. A detailed investigation of interactions within the shielding to HPGe detector response using MCNP code
  15. Cosmic ray angular distribution employing plastic scintillation detectors and flash-ADC/FPGA-based readout systems
https://doi.org/10.3139/124.110280

Articles in the same Issue

  1. Contents/Inhalt
  2. Contents
  3. Summaries/Kurzfassungen
  4. Summaries
  5. Technical Contributions/Fachbeiträge
  6. Remarks on boiling water reactor stability analysis – part 2: stability monitoring
  7. Irradiation tests on PHWR type fuel elements in TRIGA research reactor of INR Pitesti
  8. Heat transfer study of a submerged reactor channel under boil-off condition
  9. Dynamic assessment for life extension of nuclear power plants (NPPs) using system dynamics (SD) method
  10. Effect of using FLiBe and FLiNaBe molten salts bearing plutonium fluorides on the neutronic performance of PACER
  11. Evaluations of the CCFL and critical flow models in TRACE for PWR LBLOCA analysis
  12. Development of program DETSIM to simulate detector's full energy peak efficiency
  13. The criticality calculations for one-speed neutrons in a reflected slab with anisotropic scattering using the modified UN method
  14. A detailed investigation of interactions within the shielding to HPGe detector response using MCNP code
  15. Cosmic ray angular distribution employing plastic scintillation detectors and flash-ADC/FPGA-based readout systems
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