National Research Nuclear University MEPhI, Moscow, Russia:

A. B. Kruglov, Cand. Eng., Associate Professor
M. D. Savelyev, Engineer, e-mail: SavelyevMD@gmail.com
B. A. Tarasov, Cand. Eng., Senior Researcher

Pressurized water reactors (PWR, VVER, EPR, CAP) are currently the backbone of the world's nuclear power industry. The fuel in this type of power plants is uranium dioxide pellets enclosed in a sealed zirconium shell. Loss Of Coolant Accident (LOCA) emergencies can cause partial or complete destruction of zirconium claddings and accumulation of hydrogen in the reactor pressure vessel and reactor building. This is due to the so-called steam-zirconium reaction: an exothermic process of zirconium oxidation in an atmosphere of saturated water vapor at temperatures above 900 °C. The products of this reaction are hydrogen and brittle zirconium oxide, which collapses under the action of its own weight and the weight of the fuel column. Fuel for pressurized water nuclear power reactors that is resistant to a loss of coolant accident is the subject of research around the world. One of the ways of implementation is the use of alloys of the Fe – Cr – Al system as a fuel element cladding material, since they have high corrosion resistance in water. The work is devoted to the determination of the coefficient of linear thermal expansion (CLTE), specific heat capacity, thermal conductivity and the influence of composition on them for Fe – Cr – Al – Si alloys in the temperature range of 200–1000 °C. New data on the dependence of the thermophysical characteristics of alloys of the Fe – Cr – Al – Si system on temperature and chromium concentration have been obtained. It is shown that the CLTE, specific heat capacity and thermal conductivity weakly depend on the chromium content in the studied concentration range, the main contribution of alloying elements is made by the position of the Curie.

1. Wu X., Kozlowski T., Hales J. D. Neutronics and fuel performance evaluation of accident tolerant FeCrAl cladding under normal operation conditions. Annals of Nuclear Energy. 2015. Vol. 85. pp. 763–775.
2. Gamble K. A. et al. An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions. Journal of Nuclear Materials. 2017. Vol. 491. pp. 55–66.
3. Nikitin S. N. et al. Influence of a doping by Al stainless steel on kinetics and character of interaction with the metallic nuclear fuel. IOP Conference Series: Materials Science and Engineering. IOP Publishing. 2016. Vol. 130, Iss. 1. p. 012024.
4. Nikitin S. N. et al. Increasing of compatibility of metallic nuclear fuel with various construction materials. Tsvetnye Metally. 2015. No. 3. pp. 36–39.
5. Tarasov B. A., Savelyev M. D., Shornikov D. P. Corrosion Resistance of Fe – Cr – Al – Si Alloys with Low Chromium Content. KnE Materials Science. 2018. Vol. 4, Iss. 1. pp. 480–490.
6. Tarasov B. A. et al. Short-term mechanical properties of Fe – Cr – Al – Si alloys. KnE Materials Science. 2018. Vol. 4, Iss. 1. pp. 491–497.
7. Field K. G. et al. Handbook on the material properties of FeCrAl alloys for nuclear power production applications (FY18 Version: Revision 1). Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States), 2018. No. ORNL/SPR-2018/905.
8. Kim H. et al. 400 °С aging embrittlement of FeCrAl alloys: Microstructure and fracture behavior. Materials Science and Engineering: A. 2019. Vol. 743. pp. 159–167.
9. Kobayashi S., Fukunishi H. Reduction of thermal expansion of ferritic/martensitic heat resistant steels-alloying effects on thermal expansion of α-Fe phase. ISIJ International. 2020. Vol. 60, Iss. 12. pp. 2983–2989.
10. Xiong W. et al. An improved thermodynamic modeling of the Fe–Cr system down to zero kelvin coupled with key experiments. Calphad. 2011. Vol. 35, Iss. 3. pp. 355–366.
11. Kawahara K. et al. High temperature deformation and fracture in ferromagnetic Fe – Co and Fe-Cr alloys (Brittle Fracture). Proceedings of the Asian Pacific Conference on Fracture and Strength and International Conference on Advanced Technology in Experimental Mechanics 1.01. 203. The Japan Society of Mechanical Engineers, 2001. pp. 134–139.
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