Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT) (방사성폐기물학회지)

Korean Radioactive Waste Society (한국방사성폐기물학회)
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Effect of Steady-State Oxidation on Tensile Failure of Zircaloy Cladding

  • Received : 2022.03.28
  • Accepted : 2022.06.17
  • Published : 2022.06.30
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Abstract

The effect of oxidation time on the characteristics and mechanical properties of spent nuclear fuel cladding was investigated using Raman spectroscopy, tube rupture test, and tensile test. As oxidation time increased, the Raman peak associated with the tetragonal zirconium oxide phase diminished and merged with the Raman peak associated with the monoclinic zirconium oxide phase near 333 cm-1. Additionally, the other tetragonal zirconium oxide phase peak at 380 cm-1 decreased after 100 d of oxidation, whereas the zirconium monoclinic oxide peak became the dominant peak. The oxidation time had no effect on the tube rupture pressure of the oxidized zirconium alloy tube. However, the yield and tensile stresses of the oxidized nuclear fuel cladding tube decreased after 100 d of oxidation. The results of the scanning electron microscopy and transmission electron microscopy were represented with the in-situ Raman analysis result for the oxide characteristics generated on the cladding of spent nuclear fuel.

Keywords

Acknowledgement

This work was supported by the Nuclear Safety Research Program through the Korea Foundation Of Nuclear Safety (KoFONS) using the financial resource granted by the Nuclear Safety and Security Commission (NSSC) of the Republic of Korea (No. 2106022), and the KAERI Institutional Project (Project No. 522320-22).

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