The results of experimental and theoretical investigations of the corrosion of outer protective coatings, made of silicon carbide and pyrolytic carbon, of fuel microspherules in high-pressure water and water vapor in the range 350-950°C are presented. The tests lasted for times ranging from 300 to 11800 h were conducted for the operating conditions of light-water reactors. A comparative estimate is given of the corrosion resistance of two types of coatings. The temperature-time curves, which make it possible to determine by interpolation and extrapolation the degree of corrosion of the protective coatings for different operating regimes of the reactor, are obtained.In the last few years, there has been a great deal of interest in developing light-water reactors with high and supercritical steam pressure [1][2][3]. The most important factor in improving nuclear power plants is guaranteeing their safety under normal operating conditions and in emergency situations, including sabotage and terrorist acts. In addition, the problems arising from the unavoidable service life extension of operating nuclear power plants with VVÉR and RBMK type light-water reactors while maintaining high safety and reliability remain urgent.The experimental and theoretical investigations conducted for more than 10 years [4,5] show that the problems posed can be solved by using in the reactor core fuel micropellets which are indirectly cooled by coolants, in contrast to the conventional fuel pellets which are enclosed in zirconium cladding. A strong argument in favor of such a technical solution is the domestic and foreign experience in using spherical fuel elements in high-temperature helium-cooled reactors [6,7].When using fuel micropellets, one of the main problems which need to be studied is the necessity of ensuring high corrosion resistance and the integrity of the outer protective coatings, which are the main protective barrier in the path of radioactive gases from fission products in the coolant under conditions of prolonged operation of the reactor in normal regimes and during serious accidents.
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