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Conference Papers Year : 2016

CFD Analysis of a Steam Generator Separation Test in the Kozloduy VVER-1000 Reactor

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Abstract

Computational fluid dynamics (CFD) research for nuclear reactor safety dedicates to real scale reactor circuits under realistic thermal hydraulic conditions. In the framework of an OECD/NEA benchmark, CEA has attempted 10 years ago with the code TrioCFD to study the temperature distribution at the core inlet in a main steam line break (MSLB) accident scenario in a Bulgarian VVER1000 reactor. This work is resumed here by completing the geometry of the reactor pressure vessel (RPV) and by capitalizing both code development and high performance computing (HPC) resources. Before modelling the full scale RPV thermal-hydraulics, a PIRT (Phenomena Identification and Ranking Table) was performed to classify the existing physical phenomena in a ranking table. Three single effect validation test cases were defined in a test matrix. The CFD approach was validated single effect by single effect by reproducing the defined well suited test cases. The core outlet temperature distribution was measured during a commissioning steam generator separation test at Kozloduy nuclear power plant. This temperature distribution is compared to the CFD calculations and helps to validate integrally the full scale reactor calculation. Tetrahedral meshes of 50 to 400 million velocity control volumes were generated for the complete RPV; self-evidently the mesh refinement reflects the restrictions of the former defined test matrix. In the OECD benchmark, the core inlet temperature was calculated from the measured core outlet temperature by simple energy conservation. With the integral calculation we were able to review this process with the calculated core inlet and outlet temperature.
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cea-02438706 , version 1 (27-02-2020)

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  • HAL Id : cea-02438706 , version 1

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Q. Feng, U. Bieder. CFD Analysis of a Steam Generator Separation Test in the Kozloduy VVER-1000 Reactor. NUTHOS-11 The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, Oct 2016, Gyeongju, South Korea. ⟨cea-02438706⟩

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