DOI QR코드

DOI QR Code

Hydrogen Effect on the Oxidation of Zr-Alloy Claddings under High Temperature

수소화물에 의한 Zr 합금의 고온산화 가속효과

  • Jung, Yunmock (Korea Atomic Energy Research Institute) ;
  • Ha, Sungwoo (Department of Nuclear Engineering, Kyunghee University) ;
  • Park, Kwangheon (Department of Nuclear Engineering, Kyunghee University)
  • Received : 2015.11.16
  • Accepted : 2016.07.29
  • Published : 2016.08.31

Abstract

The operation method of nuclear power plants is currently changing to high burn-up and long period that can enhance economics and efficiency of the plant. Since nuclear plant operation environment has been becoming severe, the amount of absorbed hydrogen also has increased. Absorbed hydrogen can be fatal securing safety of nuclear fuel cladding in case of Loss of Coolant Accidents(LOCA). In order to examine the impact of hydride on high-temperature oxidation, high-temperature oxidation experiment was performed on normal Zry-4 cladding and on Zry-4 cladding where hydrogen is charged in air pressure steam atmosphere under the $950^{\circ}C$ and $1000^{\circ}C$. According to the results, while oxidation acceleration due to charged hydrogen was not observed prior to breakaway oxidation creation, oxidation began to accelerate in cladding where hydrogens charged as soon as the breakaway oxidation started. If so much hydrogen are charged in the cladding, equiaxial monoclinic phase to unstable of stress is formed and it is presumed that oxidation is accelerated because nearby stress caused a crack in equiaxial phase, and that makes corrosion resistance decline sharply.

Keywords

References

  1. D. D. Lanning, C. E. Beyer, and K. J. Geelhood, FRAPCON-3 Updates, Including Mixed-Oxide Fuel Properties, NRC, NUREG/CR-6534, Vol. 4 (2005), 53-57.
  2. R. E. Pawel, J. V. Cathcart, and J. J. Campbell, "The Oxidation of Zry-4 at 900 and 1100oC in High Pressure Steam", J. Nucl. Mater. 82,(1979), 129-139. https://doi.org/10.1016/0022-3115(79)90045-X
  3. Nuclear Regulatory Commission, Performance- Based Emergency Core Cooling Systems Cladding Acceptance Criteria :Proposed Rule, 10 CFR Parts 50 and 52, Federal Register. Vol.79 (2014), No.56, 34-46.
  4. ASTM, Standard Test Method for Corrosion Testing of Products of Zirconium, Hafnium, and Their Alloys in 633 K or in Steam at 673 K [Metric], Annual Books of ASTM Standard section, ASTM (1991), 3, 49.
  5. Y. S. Kim et al., A Manual for Characterization Tests for Zr-2.5Nb Pressure Tubes, KAERI Technical Report, KAERI/TR-1329/99 (1999), 3-11.
  6. A. D. Lepage, W. A. Ferris, G. A. Ledoux, Procedure for Adding Hydrogen to Small Sections of Zirconium Alloys, Materials and Mechanics Branch, Chalk River Laboratories, Chalk River, Ontario (1998), 12-13.
  7. Jong Hyuk Baek, and Yong Hwan Jeong, Breakaway Phenomenon of Zr-based Alloys during a High-Temperature Oxidation, Journal of Nuclear Materials, Vol. 372 (2008), 152-159. https://doi.org/10.1016/j.jnucmat.2007.02.011
  8. Seonggi Jeong, The Effect of Hydrogen in the Nuclear Fuel Cladding on the Oxidation under High Temperature and High Pressure Steam, Paper of masters degree, Kyunghee University (2013).
  9. Nuclear Regulatory Commission, Performance- Based Emergency Core Cooling Systems Cladding Acceptance Criteria; Proposed Rule, Federal Register, Vol. 79 (2014), No.56.
  10. Nuclear Regulatory Commission, Cladding Embrittlement during Postulated Loss-of-Coolant Accidents, NUREG/CR-6967 (2008).
  11. S. Leistikow and G.Schanz, Oxidation kinetics and related phenomena of Zry-4 fuel cladding exposed to high temperature steam and hydrogen-steam mixtures under PWR accident conditions, Nuclear Engineering and Design, Vol.103 (1987), 65-84. https://doi.org/10.1016/0029-5493(87)90286-X
  12. K. W. Lee, The influence of surface condition on the Breakaway phenomenon of Zr-alloy, Paper of masters degree, Chungnam National University (2011), 48-63.