Open Access Peer-reviewed Research Article

Main Article Content

Nadezhda Ishchenko
Ivan Petelguzov corresponding author
Olena Slabospitska


The subject of this study is the oxidation of fuel rod cladding made of material Zr1Nb(0.1% O) in steam at temperatures in the range of 660°C to 1200°C with a surface in the initial state (after manufacturing - grinding) and after additional chemical etching. The changes in the microstructure of tubes due to the interaction with steam were investigated. A comparison was made between the oxidation rate of this material (weight gain) and the data on the oxidation of other alloys for nuclear power plants. The oxidation rate of Zr1Nb(0.1% O) is close to the oxidation rate of other zirconium alloys. It is shown that after chemical treatment of the surface of the samples there is a more even growth of oxide films, and they have a smaller thickness for the same time of exposure than after mechanical grinding. Surface treatment before oxidation also affects the change of microstructure of samples when heated to high temperatures.

high-temperature oxidation, nuclear power plants, zirconium tubes, fuel rod cladding, steam, surface treatment, alloy and oxide structure, accident overheating

Article Details

How to Cite
Ishchenko, N., Petelguzov, I., & Slabospitska, O. (2019). Investigation of the interaction of material of fuel cladding for WWER-1000 reactor with steam at a temperature of accident overheatings. Materials Engineering Research, 1(2), 32-39.


  1. Vrtilkova V, Molin L, Kloc K, et al. Oxidation kinetics of Zrl%Nb fuel claddings in steam at 600-1200∘C (in Russian). Problems of Atomic Science and Technology. Series “Atomic Materials Science”, 1988, 2(27): 84-88.
  2. Institut für Material- und Festkörperforschung, Projekt Nukleare Sicherheit: Comparison of High Temperature Steam Oxidation Behavior of Zircaloy-4 versus Austenitic and Ferritic Steels under Light Water Reactor Safety Aspects. KfK 3994, Karlsruhe, 1985.
  3. Brachet JC, Pelchat J, Hamon D, et al. Mechanical behaviour at room temperature and metallurgical study of low-tin Zy-4 and M5TM after oxidation at 1100∘C and quenching. Fuel behaviour under transient and LOCA conditions: Proceedings of a Technical Committee meeting held by IAEA in Halden, Norway, 2001.
  4. Baek JH, Park KB and Yeong YH. Oxidation kinetics of Zircaloy-4 and Zr1Nb1Sn0,1Fe at temperatures of 700-1200∘C. Journal of Nuclear Materials, 2004, 335: 443-456.
  5. Petelguzov IA. The kinetics and corrosion mechanism of alloy Zr1Nb at heating in water vapour at temperature 660 to 1200∘C. Problems of Atomic Science and Technology. Series: “Physics of Radiation Effect and Radiation Materials Science”, 2006, 4(89): 97-103.
  6. Baek JH and Jeong YH. Breakaway phenomenon of Zr-based alloys during a high-temperature oxidation. Journal of Nuclear Materials, 2008, 372: 15-159.
  7. Yan Y, Burtseva TA and Billone MC. High-temperature steam oxidation behavior of Zr1Nb cladding alloy E110. Journal of Nuclear Materials, 2010, 393: 433-448.
  8. Banerjee S, Sawarn TK, Alur VD, et al. High temperature steam oxidation study on Zr2.5%Nb pressure tube under simulated LOCA condition. Journal of Nuclear Materials, 2013, 439: 258-267.
  9. Waeckel N and Mardon JP. Recent data on M5TM Alloy under LOCA Conditions (as compared to Zy-4 behavior): Nuclear Safety Research Conference (NSRC) Fuel Session 2003, Washington DC, USA, 2003.
  10. Chung HM. Fuel behavior under loss-of-coolant accident situations. Nuclear Engineering and Technology, 2005, 37: 327-362.
  11. Azhazha VM, V’yugov PN, Petelguzov IA, et al. Study of corrosion stability sample tapes from calcium-thermal alloy Zr1Nb with contents of the oxygen before 0,1% weight. Problems of Atomic Science and Technology. Series “Physics of Radiation Effect and Radiation Materials Science”, 2007, 6(91): 40-45.
  12. Krasnorutskyy VS, Petelguzov IA, Grytsyna VM, et al. Express and longtime tests of fuel pipes from alloy Zr1Nb (0.1% O). Problems of Atomic Science and Technology. Series “Pure Materials and the Vacuum Technologies”, 2011, 6(76): 42-47.
  13. Waterside Corrosion of Zirconium Alloys in Nuclear Power Plants. International Atomic Energy Agency, Vienna, 1998.
  14. Motta AT, Gomes da Silva MJ, Yilmazbayhan A, et al. Microstructure and Growth Mechanism of Oxide Layers, formed on Zr Alloys, Studied with Micro-Beam Synchrotron Irradiation. Journal of ASTM International, 2008, 5: JAI12375.
  15. Ni N, Lozano-Perez S, Jenkins ML, et al. Porosity in oxides on zirconium fuel cladding alloys, and its importance in controlling oxidation rates. Scripta Materialia, 2010, 62: 564-567.
  16. Zhou BX, Li Q, Yao MY, et al. 2008, Effect of Water Chemistry and Composition on Microstructural Evolution of Oxide on Zr Alloys, Zirconium in Nuclear industry: 15th International Symposium, ASTM International, West Consholocken, PA. ASTM STP 1505: 360-383.
  17. Gosset D, Le Saux M, Simeone D, et al. New insights in structural characterization of zirconium alloys oxidation at high temperature. Journal of Nuclear Materials, 2012, 429: 19-24.
  18. Akhiani H and Szpunar JA. Effect of surface roughness on the texture and oxidation behavior of Zircaloy-4 cladding tube. Applied Surface Science, 2013, 285: 83-839.