A variant of experimental dispersion fuel elements with fuel based on uranium dioxide as fissile material in a silumin matrix, which have been proposed for modernizing the core of the SM reactor and are irradiated in a wide spectrum of neutron-physical and thermophysical parameters in the reflector channel of the reactor, is investigated. A nonuniformity with fuel density two times higher than the average over a fuel element, viz., a cluster of uranium dioxide particles in the central zone of the kernel, has been found in the distribution of the fuel phase in the fuel elements. The thermophysical calculations show that for such a distribution of fuel particles the temperature of the fuel composition with heat flux density from the fuel-element surface >6 MW/m2 can reach the melting temperature of the matrix. This is confirmed by investigations of fuel elements with appreciably larger volume and shape change. To increase the serviceability of the fuel elements, ways of improving their design and the material composition of the fuel mix are proposed.
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References
V. A. Tsykanov, A. V. Klinov, V. A. Starkov, et al., “Modernization of the core of a SM reactor to solve materials engineering problems,” At. Énerg., 93, No. 3, 167–172 (2002).
V. A. Tsykanov, A. V. Klinov, V. A. Starkov, et al., “Main results of the first stage of modernization of the SM core,” At. Énerg., 102, No. 2, 86–92 (2007).
V. S. Volkov, A. V. Morozov, A. V. Kozlov, et al., “Development of a fuel element with low harmful absorption of neutrons for the high-flux SM research reactor,” At. Énerg., 106, No. 6, 314–318 (2009).
A. V. Klinov, N. K. Kalinina, V. V. Pimenev, et al., “Tests of experimental fuel assemblies with low harmful absorption of neutrons in the SM reactor,” Izv. Vys. Ucheb. Zaved. Yad. Energet., No. 2, 114–122 (2013).
V. D. Grachev, “Some questions concerning the mathematical implementation of the fi nite-element method in problems of reactor thermal physics,” Preprint NIIAR-6 (652), TSNIIatominform, Moscow (1985).
A. G. Samoilov, A. I. Kashtanov, and V. S. Volkov, Dispersion Fuel Elements of Nuclear Reactors, Atomizdat, Moscow (1965).
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Translated from Atomnaya Énergiya, Vol. 118, No. 2, pp. 80–84, February, 2015.
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Gil’mutdinov, I.F., Shishin, V.Y., Starkov, V.A. et al. Investigation of Experimental Dispersion Fuel Elements with an Aluminum-Based Matrix for High-Flux Research Reactors. At Energy 118, 101–107 (2015). https://doi.org/10.1007/s10512-015-9963-z
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DOI: https://doi.org/10.1007/s10512-015-9963-z