Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

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This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at ORNL. Damage propagation is postulated to occur from thermal conduction between dmaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur beause of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A parametric study was done for several uncertain variables. The study included investigating effects of plate contact area, convective ... continued below

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43 p.

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Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S. & Georgevich, V. December 31, 1995.

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Description

This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at ORNL. Damage propagation is postulated to occur from thermal conduction between dmaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur beause of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A parametric study was done for several uncertain variables. The study included investigating effects of plate contact area, convective heat transfer coefficient, thermal conductivity on fuel swelling, and initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects of damage propagation. Results provide useful insights into how variouss uncertain parameters affect damage propagation.

Physical Description

43 p.

Notes

INIS; OSTI as DE96008578

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  • 7. international topical meeting on nuclear reactor thermal-hydraulics (Nureth-7), Saratoga Springs, NY (United States), 10-15 Sep 1995

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  • Other: DE96008578
  • Report No.: CONF-950904--8
  • Grant Number: AC05-84OR21400
  • DOI: 10.2172/168359 | External Link
  • Office of Scientific & Technical Information Report Number: 220557
  • Archival Resource Key: ark:/67531/metadc672943

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  • December 31, 1995

Added to The UNT Digital Library

  • June 29, 2015, 9:42 p.m.

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  • Jan. 22, 2016, 12:13 p.m.

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Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S. & Georgevich, V. Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source reactor at Oak Ridge National Laboratory, article, December 31, 1995; Tennessee. (digital.library.unt.edu/ark:/67531/metadc672943/: accessed August 19, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.