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Simulating the Thermal Interaction between Fuel and Sodium Coolant Using the EUCLID/V2 Integrated Code

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Abstract—

This article addresses the development of approaches to numerically analyzing the processes of interaction between liquid metal sodium coolant and destructed fuel-pin components (fuel and steel in solid and liquid states). Such processes may occur during a severe accident involving core destruction, and also when fuel-pin components (fuel or cladding) heated to a high temperature release into the flow of relatively cold liquid coolant or when the molten fuel begins to melt the corium catcher. A dramatic growth of power caused by self motion of pins and stoppage of forced coolant circulation without actuation of the reactor plant’s active and passive safety systems are among possible events leading to accidents with such consequences. For simulating the thermal interaction, it is proposed to use a multicomponent thermally nonequilibrium model based on the solution of a system of mass, energy, and momentum conservation equations with the relevant relationships that take into account the specific features of thermal and mechanical interaction between the melt and coolant. Simulation of the processes is very important for determining pressure jumps in the reactor plant caused by release of destructed fuel-pin components into the coolant flow. Thermal interaction of fuel-pin components with the coolant may cause intense coolant evaporation and, as a consequence, the occurrence of drastic pressure jumps determined by the intensity of heat transfer from components to coolant and the amount of vapor produced. To find the rate of heat transfer between various components, a chart of heat-transfer modes and closing relationships corresponding to each mode are used.

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Funding

This study was carried out within the framework of state contract no. N.4o.241.19.21.1068 dated April 14, 2021 for carrying out research works “Development of Integrated New-Generation Systems of Codes for Development of Nuclear Reactors and Their Safety Substantiation, 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

    E. V. Usov, V. I. Chukhno, I. A. Klimonov, A. A. Butov, N. A. Mosunova & V. F. Strizhov

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  1. E. V. Usov
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  2. V. I. Chukhno
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  3. I. A. Klimonov
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  4. A. A. Butov
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  5. N. A. Mosunova
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  6. V. F. Strizhov
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Correspondence to E. V. Usov.

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Translated by V. Filatov

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Usov, E.V., Chukhno, V.I., Klimonov, I.A. et al. Simulating the Thermal Interaction between Fuel and Sodium Coolant Using the EUCLID/V2 Integrated Code. Therm. Eng. 69, 838–843 (2022). https://doi.org/10.1134/S004060152211009X

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  • DOI: https://doi.org/10.1134/S004060152211009X

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