| Autor: |
Sergey M. Dmitriyev, Anton V. Gerasimov, Aleksander A. Dobrov, Denis V. Doronkov, Aleksey N. Pronin, Anton V. Ryazanov, Dmitry N. Solntsev, Aleksander Ye. Khrobostov, Aleksey S. Noskov, Oleg B. Samoylov, Yury K. Shvetsov, Dmitry L. Shipov |
| Jazyk: |
angličtina |
| Rok vydání: |
2021 |
| Předmět: |
|
| Zdroj: |
Nuclear Energy and Technology, Vol 7, Iss 1, Pp 49-54 (2021) |
| Druh dokumentu: |
article |
| ISSN: |
2452-3038 |
| DOI: |
10.3897/nucet.7.65313 |
| Popis: |
The paper presents the results of an experimental study to investigate the coolant interaction in adjoining fuel assemblies in the VVER reactor core composed of TVSA-T and upgraded TVSA FAs. The processes of the in-core coolant flow were simulated in a test wind tunnel. The experiments were conducted using models representing different portions of the VVER reactor core fuel bundle and consisted in measuring the radial and axial airflow velocities in representative areas within the FAs and in the interassembly space. The results of the experiments can be translated to the full-scale conditions of the coolant flow with the use of the fluid dynamics simulation theory. The measurements were performed using a five-channel pressure-tube probe. The coolant flow pattern in different portions of the fuel bundle is represented by distribution diagrams and distribution maps for the radial and axial velocity vector components in the representative areas of the models. An analysis for the spatial distribution of the radial and axial velocity vector components has made it possible to obtain a detailed pattern of the coolant flow about the FA spacer, mixing and combined spacer grids of different designs. The accumulated database for the coolant flow in FAs of different designs forms the basis for the engineering justification of the VVER reactor core reliability and serviceability. The investigation results for the coolant interaction in adjoining TVSA FAs of different designs have been adopted for the practical use at JSC Afrikantov OKBM to estimate the heat-engineering reliability of the VVER reactor cores and have been included in the database for verification of computational fluid dynamics (CFD) codes and detailed by-channel calculation codes. |
| Databáze: |
Directory of Open Access Journals |
| Externí odkaz: |
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