Open Access Peer-reviewed Research Article

Investigation of the interaction of material of fuel cladding for WWER-1000 reactor with steam at a temperature of accident overheatings

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.


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