Numerical modeling of two-phase underexpanded reactive CO$_2$-into-sodium jets in the frame ofSodium nuclear Fast Reactors - CEA - Commissariat à l’énergie atomique et aux énergies alternatives Accéder directement au contenu
Communication Dans Un Congrès Année : 2015

Numerical modeling of two-phase underexpanded reactive CO$_2$-into-sodium jets in the frame ofSodium nuclear Fast Reactors

Résumé

Supercritical CO$_2$ (sCO$_2$) Brayton cycles have gained interest in the frame of Sodium-cooled nuclear Fast Reactors (SFRs), as an alternative to the conventional water Rankine cycles. If CO$_2$ leaks inside the CO$_2$- Na heat exchanger, an underexpanded CO$_2$-into-liquid-sodium jet is formed. CO$_2$ leaks at sonic velocity and chemically reacts with sodium, through an exothermic reaction. The consequences of such a scenario must be investigated, in order to predict the temperature increasing inside the heat exchanger and on the tube walls, due to the exothermic chemical reaction, as well as the reaction products distribution inside the heat exchanger. This article presents a numerical approach for modeling such a two-phase reactive jet. A two-fluid multi-component CFD approach is employed, with a heterogeneous reaction between the CO$_2$-gas and the sodium-liquid phases. The model allows to predict the most relevant information, such as temperature distribution, the jet penetration length and the reaction products distribution downstream the CO$_2$ leakage. Some experimental studies on underexpanded gas-into-sodium reactive jets, available in literature, have been compared to our numerical results. It is found that the numerical temperature profiles are consistent with the ones experimentally measured.
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Dates et versions

cea-02509164 , version 1 (16-03-2020)

Identifiants

  • HAL Id : cea-02509164 , version 1

Citer

D. Vivaldi, F. Gruy, C. Perrais. Numerical modeling of two-phase underexpanded reactive CO$_2$-into-sodium jets in the frame ofSodium nuclear Fast Reactors. NURETH 16 - 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Aug 2015, Chicago, United States. ⟨cea-02509164⟩
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