Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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PROPERTIES OF ZR ALLOY CLADDING AFTER SIMULATED LOCA OXIDATION AND WATER QUENCHING

  • Kim, Hyun-Gil (Nuclear Convergence Technology Division, Korea Atomic Energy Research Institute) ;
  • Kim, Il-Hyun (Nuclear Convergence Technology Division, Korea Atomic Energy Research Institute) ;
  • Jung, Yang-Il (Nuclear Convergence Technology Division, Korea Atomic Energy Research Institute) ;
  • Park, Jeong-Yong (Nuclear Convergence Technology Division, Korea Atomic Energy Research Institute) ;
  • Jeong, Yong-Hwan (Nuclear Convergence Technology Division, Korea Atomic Energy Research Institute)
  • Received : 2009.07.29
  • Accepted : 2010.03.15
  • Published : 2010.04.30
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Abstract

In order to study the cladding properties of zirconium after a loss-of-coolant accident (LOCA)-simulation oxidation and water quenching test, commercial Zircaloy-4 and two kinds of HANA claddings were oxidized at temperatures ranging from $900^{\circ}C$ to $1250^{\circ}C$ and exposed for 300 s, and then cooled to $700^{\circ}C$ before quenching. Microstructural observations were made to evaluate the matrix characteristics with the chemical compositions after the LOCA-simulation test. Ring compression testing was then performed to compare the ductile behaviour of the HANA and Zircaloy-4 claddings. An X-ray diffraction (XRD) analysis was carried out for temperatures ranging from room temperature to $1250^{\circ}C$ for the oxide layer to verify the oxide crystal structure at each oxidation temperature.

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References

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