The results of calculations of the equilibrium isotopic composition of molten-salt fuel in an experimental channel of the MBIR reactor are presented. Two organizational variants of the fuel cycle for a molten-salt test loop are examined: with extraction of protactinium and without extraction of heavy nuclei. The results show that the maximum thermal power of the test loop channel can reach 12 MW. Reprocessing of the fuel salt at the rate 27.1 kg/day is necessary to maintain the thermal power at the level 1 MW. A brief description is given of the procedure for obtaining an equilibrium isotopic composition in the ISTAR software system.
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References
E. P. Velikhov, E. A. Azizov, P. N. Alekseev, et al., “The concept of ‘green’ nuclear energy,” Vopr. At. Nauki Tekhn. Ser. Termoyad. Sintez, 36, No. 1, 5–16 (2013).
Yu. G. Dragunov, I. T. Tretyakov, A. V. Lopatkin, et al., “Multipurpose Fast Research Reactor (MBIR) – innovative tool for the development of nuclear power technologies,” At. Energ., 113, No. 1, 25–28 (2012); Atomic Energy, 113, No. 1, 24–28 (2012).
S. A. Subbotin, V. V. Efremov, and V. Yu. Blandinskiy, “Forecasting system requirements to the materials of the shell of fuel elements of innovative fast reactors,” KnE Mater. Sci., 280–286 (2018).
V. M. Azhazha, “Alloys for molten-salt reactors,” Vopr. At. Nauki Tekhn. Ser. Fiz. Radiats. Povrezhd. Radiats. Materialoved., No. 4, 40 –47 (2005).
A. Rykhlevskii, J. Bae, and K. Huff, “Modeling and simulation of online reprocessing in the thorium-fueled molten salt breeder reactor,” Ann. Nucl. Energy, 128, 366–379 (2019).
E. Capelli, “Thermodynamic investigation of the LiF–ThF4 system,” Chem. Thermodyn., 58, 110–116 (2013).
B. Porter and R. Meaker, Density and Molar Volumes of Binary Fluoride Mixtures, US Dept. of the Interior, Bureau of Mines (1966).
C. Weaver, “Phase equilibria in the systems UF4–ThF4 and LiF–UF4–ThF4,” J. Am. Ceramic Soc., 43, No. 4, 213–218 (1960).
L. D. Alekseevskii, “Search for a possible structure of a stationary system of future nuclear power with a closed nuclear fuel cycle based on the study of nuclide balances,” Vopr. At. Nauki Tekhn. Ser. Fiz. Yad. Reakt., No. 2, 21–26 (2008).
V. Yu. Blandinskiy and A. A. Dudnikov, “Calculations of spent fuel isotopic composition for fuel rod from VVER-440 fuel assembly benchmark using several evaluated nuclear data libraries,” Kerntechnik, 83, No. 4, 325–330 (2018).
V. M. Novikov, I. S. Slesarev, P. N. Alekseev, et al., Nuclear Reactors with Enhanced Safety (analysis of conceptual developments), Energoatomizdat, Moscow (1993).
MCNP – a General Monte Carlo N-Particle Transport Code, Version 5/X-5 Monte Carlo Team. LA-UR-03-1987.
V. L. Blinkin and V. M. Novikov, Molten Salt Nuclear Reactors, Atomizdat, Moscow (1978).
V. V. Ignatiev, O. S. Feinberg, A. V. Zagnitko, et al., “Molten-salt reactors: new possibilities, problems and solutions,” At. Energ., 112, No. 3, 135–143 (2012); Atomic Energy, 112, No. 3, 157–165 (2012).
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Translated from Atomnaya Énergiya, Vol. 128, No. 5, pp. 254–258, May, 2020.
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Blandinskii, V.Y., Kuzenkova, D.S. Computational Validation of Experiments with Molten-Salt Thorium-Uranium Fuel Compositions in MBIR Test Loop. At Energy 128, 271–276 (2020). https://doi.org/10.1007/s10512-020-00687-3
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DOI: https://doi.org/10.1007/s10512-020-00687-3