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Simulation of Melt Behavior in the Sodium-Cooled Reactor Core Catcher Using the EUCLID/V2 Integrated Computer Code HEFEST-FR Module

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Abstract—

For numerically simulating the melt behavior in the core catcher of a sodium-cooled reactor, the HEFEST-FR module—a software tool based on the SAFR computer code elaborated at the Nuclear Safety Institute, Russian Academy of Sciences, for simulating the meltdown and destruction of liquid metal-cooled fast reactor core components—has been developed and incorporated into the EUCLID/V2 integrated computer code. This module is intended for numerically simulating the melt retention and cooling-down processes in the reactor vessel with taking into account the heat transfer from the vessel’s internal structures to the coolant. For this purpose, a 2D-problem (implying that there is no dependence of temperature on the azimuthal angle) of unsteady heat conduction for materials located in the reactor’s core catcher tray is solved in a cylindrical coordinate system. The heat-conduction equation coefficients depend on time, coordinates, and temperature, the latter being the solution of equation. Boundary conditions of the first, second, and third kind are used, and heat loss by radiation at the boundary is specified. The total or volumetric decay heat power is given inside of the melt. For numerically solving the 2D heat-conductivity problem, an enthalpy approach based method is applied. The formulations used in the method make it possible to overcome, in a natural manner, the problem relating to discontinuity of material specific melting point enthalpy in solving the heat-conduction problem with taking phase transitions into account. The solution yields the temperature field in melting/freezing of materials (steels of various grades and fuel) in the fast reactor’s core catcher tray. The results of verifying the HEFEST-FR module against the solution of an analytical problem have been demonstrated. Using the module, methodical computations of the fuel and fuel pin cladding melt behavior in the sodium-cooled reactor’s core catcher have been carried out.

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REFERENCES

  1. A. I. Bel’tyukov, A. I. Karpenko, S. A. Poluyaktov, O. L. Tashlykov, G. P. Titov, A. M. Tuchkov, and S. E. Shcheklein, Nuclear Power Plants with Fast Neutron Reactors with Sodium Coolant: Textbook (Ural. Fed. Univ., Yekaterinburg, 2013), Vol. 1 [in Russian].

    Google Scholar 

  2. O. L. Tashlykov, Fundamentals of Nuclear Energy: Tutorial (Ural. Fed. Univ., Yekaterinburg, 2016) [in Russian].

    Google Scholar 

  3. I. A. Kuznetsov and V. M. Poplavskii, Safety of Nuclear Power Plants with Fast Neutron Reactors (IzdAt, Moscow, 2012) [in Russian].

    Google Scholar 

  4. Yu. G. Dragunov, V. V. Lemekhov, A. V. Moiseev, V. S. Smirnov, O. A. Yarmolenko, V. P. Vasyukhno, and Yu. S. Cherepnin, “Technical project RU BREST-OD-300: Stages of development and substantiation,” in Innovational Projects and Technologies of Nuclear Power Industry: Proc. 4th Int. Sci. and Tech. Conf., Moscow, Russia, Sept. 27–30, 2016 (NIKIET, Moscow, 2016), Vol. 1, pp. 21–30.

  5. E. O. Adamov, “Closed fuel cycle technologies based on fast reactors as the corner stone for sustainable development of nuclear power,” in Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable Development (FR17): Proc. Int. Conf., Yekaterinburg, Russia, June 26–29, 2017 (International Atomic Energy Agency, 2017), paper no. IAEA-CN245-342.

  6. 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 

  7. V. M. Alipchenkov, A. V. Boldyrev, D. P. Veprev, Yu. A. Zeigarnik, P. V. Kolobaeva, E. V. Moiseenko, N. A. Mosunova, E. F. Seleznev, V. F. Strizhov, E. V. Usov, S. L. Osipov, V. S. Gorbunov, D. A. Afremov, and A. A. Semchenkov, “The EUCLID/V1 integrated code for safety assessment of liquid metal cooled fast reactors. Part 2: Validation and verification,” Therm. Eng. 65, 627–640 (2018). https://doi.org/10.1134/S004060151809001X

    Article  Google Scholar 

  8. Dynamic Integral Universal Calculation Code for Safety Analysis and Substantiation of Liquid Metal Cooled Fast Neutron Reactor Plants, Version 1.2 (EUCLID/V1.2): Software Attestation Passport No. 462 (Rostekhnadzor, Moscow, 2019) [in Russian].

  9. 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 

  10. A. A. Butov, V. S. Zhdanov, I. A. Klimonov, I. G. Kudashov, A. E. Kutlimetov, P. D. Lobanov, N. A. Mosunova, A. A. Sorokin, V. F. Strizhov, 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 

  11. A. Bachrata, F. Bertrand, N. Marie, A. Edeline, R. Kubota, K. Kamiyama, and S. Kubo, “Severe accident studies on the efficiency of mitigation devices in a SFR core with SIMMER code,” Nucl. Eng. Des. 373, 111037 (2021). https://doi.org/10.1016/j.nucengdes.2020.111037

    Article  Google Scholar 

  12. A. Lecoanet, F. Payot, C. Journeau, N. Rimbert, and M. Gradeck, “Study of the ablation consecutive to jet impingement on a meltable solid — Application to SFR core-catcher,” Nucl. Eng. Des. 377, 111147 (2021). https://doi.org/10.1016/j.nucengdes.2021.111147

    Article  Google Scholar 

  13. G. Bala Sundaram and K. Velusamy, “Development of a robust multi-phase heat transfer model and optimization of multi-layer core catcher for future Indian sodium cooled fast reactors,” Ann. Nucl. Energy 136, 107042 (2020). https://doi.org/10.1016/j.anucene.2019.107042

    Article  Google Scholar 

  14. G. Bala Sundaram and K. Velusamy, “A comparative assessment of various single tray core catcher options for future sodium cooled fast reactors,” Prog. Nucl. Energy 123, 103309 (2020). https://doi.org/10.1016/j.pnucene.2020.103309

    Article  Google Scholar 

  15. G. Bala Sundaram and K. Velusamy, “Effect of debris material composition on post accidental heat removal in a sodium cooled fast reactor,” Nucl. Eng. Des. 375, 111065 (2021). https://doi.org/10.1016/j.nucengdes.2021.111065

    Article  Google Scholar 

  16. ANSYS Advantage Magazine. https://www.ansys.com

  17. A. S. Filippov, Development, Verification, and Application of Software for Calculation Analysis of the Late Stage of a Severe Accident at a Nuclear Power Plant with VVER, Doctoral Dissertation in Engineering (Nuclear Safety Inst. of the Russian Academy of Sciences, Moscow, 2013).

  18. A. S. Filippov, N. I. Drobyshevskii, A. E. Kiselev, V. F. Strizhov, and A. L. Fokin, “SOCRAT/HEFEST code: Models of VVER core melt interaction with reactor structures under severe accident conditions,” Izv. Ross. Akad. Nauk, Energ., No. 3, 4–24 (2010).

  19. V. Alexiades and A. D. Solomon, Mathematical Modeling of Melting and Freezing Processes (Routledge, New York, 1993). https://doi.org/10.1201/9780203749449

    Book  Google Scholar 

  20. A. V. Lykov, Theory of Heat Conduction (Vysshaya Shkola, Moscow, 1967) [in Russian].

    Google Scholar 

  21. G. Kayser, “The SCARABEE BF1 experiment with a molten UO2 pool and its interpretation,” in Proc. Workshop on Large Molten Pool Heat Transfer, Grenoble, France, March 9–11, 1994 (OECD Nuclear Energy Acency, 1994), pp. 207–225.

  22. S. V. Ovchinnikov, Introduction to Heat Transfer Theory: Heat Conduction in Solids: Methodical Tutorial (Sarat. Gos. Univ. im. N. G. Chernyshevskogo, Saratov, 2015) [in Russian].

  23. E. M. Kartashov, Analytical Methods in Solids Heat Conduction Theory: Tutorial (Vysshaya Shkola, Moscow, 2001) [in Russian].

    Google Scholar 

  24. R. B. Kotel’nikov, S. N. Bashlykov, A. I. Kashtanov, and T. S. Men’shikova, High-Temperature Nuclear Fuel (Atomizdat, Moscow, 1978) [in Russian].

    Google Scholar 

  25. N. M. Mitrofanova, A. V. Tselishchev, V. S. Ageev, Yu. P. Budanov, A. G. Ioltukhovskii, M. V. Leont’eva-Smirnova, F. G. Reshetnikov, Yu. K. Bibilashvili, I. A. Shkabura, and Yu. A. Ivanov, “Structural materials for fuel rod cladding and fuel assembly jackets of the BN-600 reactor,” Izv. Vyssh. Uchebn. Zaved., Yad. Energ., No. 1, 211–223 (2011).

  26. Thermophysical Properties of Materials for Nuclear Engineering. A Tutorial and Collection of Data (International Atomic Energy Agency, Vienna, 2008).

  27. J. H. Harding, D. G. Martin, and P. E. Potter, Thermophysical and Thermochemical Properties of Fast Reactor Materials (Commission of the European Communities, Harwell Laboratory, UK, 1989).

    Google Scholar 

  28. P. L. Kirillov, Thermophysical Properties of Materials for Nuclear Engineering (IzdAT, Moscow, 2007) [in Russian].

  29. V. S. Chirkin, Thermophysical Properties of Materials for Nuclear Engineering (Atomizdat, Moscow, 1968) [in Russian].

    Google Scholar 

Download references

Funding

This work was financially supported by the State Atomic Energy Corporation Rosatom within the framework of State Contract no. N.4o.241.19.21.1068 dated April 14, 2021, for carrying out scientific research works: “Development of New-Generation Systems of Integrated Computer Codes for Development of Nuclear Reactors and Substantiation of Their Safety, Designing of Nuclear Power Plants, and Development of Nuclear Fuel Cycle Technologies and Facilities. Stage of 2021–2023.”

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Authors and Affiliations

  1. Nuclear Safety Institute, Russian Academy of Sciences (IBRAE RAS), 115191, Moscow, Russia

    D. D. Kamenskaya, N. A. Mosunova & S. V. Tsaun

  2. Novosibirsk Branch, Nuclear Safety Institute, Russian Academy of Sciences, 630090, Novosibirsk, Russia

    A. A. Butov, I. A. Klimonov, E. V. Usov & V. I. Chukhno

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  1. A. A. Butov
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  2. D. D. Kamenskaya
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  3. I. A. Klimonov
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  4. N. A. Mosunova
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  6. S. V. Tsaun
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  7. V. I. Chukhno
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Correspondence to S. V. Tsaun.

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Translated by V. Filatov

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Butov, A.A., Kamenskaya, D.D., Klimonov, I.A. et al. Simulation of Melt Behavior in the Sodium-Cooled Reactor Core Catcher Using the EUCLID/V2 Integrated Computer Code HEFEST-FR Module. Therm. Eng. 70, 841–848 (2023). https://doi.org/10.1134/S0040601523110034

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