Coolant Hydrodynamics at the Inlet to the FA of the RITM-Type Reactor of a Small Nuclear Power Plant
The results of an experimental study into the features of the process of coolant flow formation in the inlet section of the fuel assembly (FA) of the core of a RITM-type reactor of a small nuclear power plant (SNPP) are presented. The purpose of the work is to evaluate the influence of different des...
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Veröffentlicht in: | Thermal engineering 2024-05, Vol.71 (5), p.375-390 |
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Sprache: | eng |
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Zusammenfassung: | The results of an experimental study into the features of the process of coolant flow formation in the inlet section of the fuel assembly (FA) of the core of a RITM-type reactor of a small nuclear power plant (SNPP) are presented. The purpose of the work is to evaluate the influence of different design elements of the inlet section on the formation of coolant flow. To achieve this goal, a series of experiments was completed on a research aerodynamic stand with an air working environment using a large-scale experimental model, including structural elements of the FA inlet section from the throttle orifice to the fuel rod fastening unit to the diffuser, as well as a fragment of the fuel rod bundle between the absorber and spacer grids. The studies were carried out using the pneumometric method and the method of injection of a contrast admixture in several sections along the length of the model. Measurements were made over the entire cross section of the model. Features of the coolant flow are visualized by cartograms of the axial flow velocity of the working medium and the distribution of admixture in the cross section of the model. The research results were used by specialists from the design and calculation departments of OKBM Afrikantov to substantiate engineering solutions when designing new cores of RITM reactors. The results of the experiments were collected into a database and used in the validation of the LOGOS CFD computer program created by employees of the All-Russian Research Institute of Experimental Physics and the Institute for Theoretical and Mathematical Physics of Moscow State University as analogues of foreign computer programs of the same class, which include ANSYS, Star CCM, etc. Experimental data were also used when validating one-dimensional thermal-hydraulic codes used at OKBM Afrikantov in substantiating the thermal reliability of reactor cores. The thermohydraulic code CANAL is also included in this class of computer programs. |
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ISSN: | 0040-6015 1555-6301 |
DOI: | 10.1134/S0040601524050057 |