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Numerical Investigation of the Molten Metallic and Oxide Fuel Relocation along the Surface of Fuel Pin

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Abstract

This paper presents the results from numerical investigation of the molten (metal or oxide) fuel relocation during a thermal failure of a fuel rod under conditions close to a severe accident with a power increase in a fast rector. Since experimental investigations needed to determine the regularities of an accident with core destruction cannot be carried out under actual reactor conditions due to safety reasons, getting information on the specifics of molten fuel’s movement using mathematical simulation methods is urgent. The methods and approaches used in the considered problem to simulate molten fuel flow and heat transfer with the fuel-rod surface are briefly described. Although the main fuel in reactor units is an oxide fuel at present, the calculations were also performed for metallic (uranium) fuel to evaluate its effect on the development of an accident. In the calculations, account was taken for the different regularities of heat transfer of metal and oxide fuels that resulted from a considerable difference in the Prandtl numbers for melts of these materials. The processes were investigated for the first phase of severe accidents covering the period from the onset of fuel-rod melting to the melt escape from the core center. Fuel-rod melting and fuel flow were simulated for a single fuel rod.

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REFERENCES

  1. A. A. Butov, V. S. Zhdanov, I. A. Klimonov, I. G. Kudashov, A. E. Kutlimetov, N. A. Mosunova, V. F. Strizhov, A. A. Sorokin, S. A. Frolov, E. V. Usov, and V. I. Chukhno, “Verification of the EUCLID/V2 code based on experiments involving destruction of a liquid metal cooled reactor’s core components,” Therm. Eng. 66, 302–309 (2019). https://doi.org/10.1134/S0040601519050033

    Article  Google Scholar 

  2. N. A. Mosunova, “The EUCLID/V1 integrated code for safety assessment of liquid metal cooled fast reactors. Part 1: Basic models,” Therm. Eng. 65, 304–316 (2018). https://doi.org/10.1134/S0040601518050063

    Article  Google Scholar 

  3. M. Epstein, “The growth and decay of a frozen layer in forced flow,” Int. J. Heat Mass Transfer 19, 1281–1288 (1976). https://doi.org/10.1016/0017-9310(76)90080-6

    Article  Google Scholar 

  4. M. Epstein and G. M. Hauser, “The melting of finite steel slabs in flowing nuclear reactor fuel,” Nucl. Eng. Des. 52, 411–428 (1979). https://doi.org/10.1016/0029-5493(79)90030-X

    Article  Google Scholar 

  5. M. Bottoni, “Calculation of temperature distribution in a melting clad with the Pekeris–Slichter series expansion method,” Nucl. Eng. Des. 43, 249–257 (1977). https://doi.org/10.1016/0029-5493(77)90003-6

    Article  Google Scholar 

  6. E. V. Usov, A. A. Butov, S. I. Lezhnin, and P. D. Lobanov, “Solving the Stefan problem in the relation to melting of fuel elements of fast nuclear reactors,” J. Eng. Thermophys. 27, 545–553 (2018). https://doi.org/10.1134/S1810232818040173

    Article  Google Scholar 

  7. G. N. Vlasichev, “Numerical simulation of the motion and solidification of melted fuel during a serious accident in a fast reactor,” At. Energy 90, 357–365 (2001). https://doi.org/10.1023/A:1011316323468

    Article  Google Scholar 

  8. M. Ishii, W. L. Chen, and M. A. Grolmes, “Molten clad motion model for fast reactor loss-of-flow accidents,” Nucl. Sci. Eng. 60, 435–451 (1976). https://doi.org/10.13182/NSE76-3

    Article  Google Scholar 

  9. M. M. Rahman, Y. Ege, K. Morita, K. Nakagawa, K. Fukuda, and W. Maschek, “Simulation of molten metal freezing behavior on to a structure,” Nucl. Eng. Des. 238, 2706–2717 (2008). https://doi.org/10.1016/j.nucengdes.2008.04.008

    Article  Google Scholar 

  10. A. Dubey and A. Sharma, “Melting and multi-phase flow modeling of nuclear fuel in fast reactor fuel rod,” Int. J. Therm. Sci. 125, 256–272 (2018).

    Article  Google Scholar 

  11. T. Sawada, H. Ninokata, and A. Shimizu, “Calculation of a material relocation experiment simulating a core disruptive accident condition in fast breeder reactors,” Nucl. Eng. Des. 157, 177–196 (1995). https://doi.org/10.1016/0029-5493(95)00978-L

    Article  Google Scholar 

  12. E. V. Usov, A. A. Butov, V. I. Chukhno, I. A. Klimonov, I. G. Kudashov, V. S. Zhdanov, N. A. Pribaturin, N. A. Mosunova, and V. F. Strizhov, “Fuel pin melting in a fast reactor and melt solidification: Simulation using the SAFR/V1 Module of the EVKLID/V2 integral code,” At. Energy 124, 147–153 (2018).

    Article  Google Scholar 

  13. E. V. Usov, A. L. Butov, V. I. Chukhno, I. A. Klimonov, I. G. Kudashov, V. S. Zhdanov, N. A. Pribaturin, N. A. Mosunova, and V. F. Strizhov, “SAFR/V1 (EVKLID/V2 integral code module) aided simulation of melt movement along the surface of a fuel element in a fast reactor during a serious accident,” At. Energy 124, 232–237 (2018).

    Article  Google Scholar 

  14. R. Doerner, T. Bauer, J. Morman, and J. Holland, “Features of postfailure fuel behavior in transient overpower and transient undercooled/overpower tests in the transient reactor test facility,” Nucl. Technol. 98, 124–136 (1992). https://doi.org/10.13182/NT92-A34656

    Article  Google Scholar 

  15. C. Dickerman, E. Sowa, J. Monaweck, and A. Barsell, “In-pile experiments on meltdown of EBR-II Mark I fuel elements in stagnant sodium,” Nucl. Sci. Eng. 18, 319–328 (1964). https://doi.org/10.13182/NSE64-A20052

    Article  Google Scholar 

  16. S. A. Wright, G. Schumacher, and P. R. Henkel, “In-pile observations of fuel and clad relocation during LMBFR core disruptive accidents in the STAR Experiments,” Nucl. Technol. 71, 187–216 (1985). https://doi.org/10.13182/NT85-A33719

    Article  Google Scholar 

  17. K. Konishi, J. Toyooka, K. Kamiyama, I. Sato, S. Kubo, S. Kotake, K. Koyama, A. D. Vurim, V. A. Gaidaichuk, A. V. Pakhnits, and Y. S. Vassiliev, “The result of a wall failure in-pile experiment under the EAGLE project,” Nucl. Eng. Des. 237, 2165–2174 (2007). https://doi.org/10.1016/j.nucengdes.2007.03.012

    Article  Google Scholar 

  18. G. B. Usynin, Yu. I. Anoshkin, and M. A. Semenychev, “Investigating fuel element melt in simulators with fuel compositions,” At. Energy 70, 136–138 (1991).

    Article  Google Scholar 

  19. G. B. Usynin, Yu. I. Anoshkin, G. N. Vlasichev, Yu. N. Galitskikh, and N. A. Semenychev, “Model study of processes accompanying overheating of fuel elements,” At. Energy 61, 906–910 (1986).

    Article  Google Scholar 

  20. A. A. Butov, V. S. Zhdanov, I. A. Klimonov, I. G. Kudashov, A. E. Kutlimetov, N. A. Mosunova, V. F. Strizhov, A. A. Sorokin, S. A. Frolov, E. V. Usov, and V. I. Chukhno, “The EUCLID/V2 code physical models for calculating fuel rod and core failures in a liquid metal cooled reactor,” Therm. Eng. 66, 293–301 (2019). https://doi.org/10.1134/S0040601519050021

    Article  Google Scholar 

  21. B. G. Ganchev, Cooling the Elements of Nuclear Reactors with Flowing Films (Energoatomizdat, Moscow, 1987) [in Russian].

    Google Scholar 

  22. G. I. Gimbutis, “Heat transfer in the flow of a liquid-metal film under gravity on a vertical wall,” J. Eng. Phys. V. 32, 115–119 (1977). https://doi.org/10.1007/BF00858492

    Article  Google Scholar 

  23. D. Butterworth and G. F. Hewitt, Two-Phase Flow and Heat Transfer (Energiya, Moscow, 1980; Oxford Univ. Press, Oxford, 1965).

  24. S. V. Alekseenko, V. E. Nakoryakov, and B. G. Pokusaev, Wave Flow of Liquid Films (Nauka, Novosibirsk, 1992) [in Russian].

    MATH  Google Scholar 

  25. V. M. Alipchenkov, A. M. Anfimov, D. A. Afremov, V. S. Gorbunov, Yu. A. Zeigarnik, A. V. Kudryavtsev, S. L. Osipov, N. A. Mosunova, V. F. Strizhov, and E. V. Usov, “Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems,” Therm. Eng. 63, 130–139 (2016). https://doi.org/10.1134/S0040601516020014

    Article  Google Scholar 

  26. E. V. Usov, A. A. Butov, G. A. Dugarov, I. G. Kudashov, S. I. Lezhnin, N. A. Mosunova, and N. A. Pribaturin, “System of closing relations of a two-fluid model for the HYDRA-IBRAE/LM/V1 code for calculation of sodium boiling in channels of power equipment,” Therm. Eng. 64, 504–510 (2017). https://doi.org/10.1134/S0040601517070102

    Article  Google Scholar 

  27. G. Berthoud and B. Duret, “The freezing of molten fuel: reflections and new results,” in Proc. 4th Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-4), Karlsrue, Germany, Oct. 10–13,1989 (Braun, Karlsruhe, 1989), Vol. 1, pp. 675–681.

  28. E. V. Usov, A. A. Butov, V. I. Chukhno, I. A. Klimonov, I. G. Kudashov, V. S. Zhdanov, N. A. Pribaturin, N. A. Mosunova, and V. F. Strizhov, “Experiment-based verification of the SAFR/V1 module of the EVKLID/V2 integral code for thermal breakdown of fuel pins in a fast reactor,” At. Energy 124, 287–291 (2018).

    Article  Google Scholar 

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Funding

The study was supported by the Russian Science Foundation (grant no. 18-79-10013).

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Correspondence to E. V. Usov.

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Translated by T. Krasnoshchekova

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Usov, E.V., Klimonov, I.A. & Butov, A.A. Numerical Investigation of the Molten Metallic and Oxide Fuel Relocation along the Surface of Fuel Pin. Therm. Eng. 67, 122–128 (2020). https://doi.org/10.1134/S0040601520020056

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