Application of a CFD analysis to the flow behaviors in a finned rod bundle
- 1. Research Reactor Engineering Division, Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)
Description
Since the heat flux of the rod type fuel used in the HANARO, a research reactor operated in the KAERI, is substantially higher than the heat flux of power reactors, the HANARO fuel has 8 longitudinal fins for enhancing the heat release from the fuel rod surface. This unique shape of a nuclear fuel led us to study the flows and thermal hydraulic characteristics of it. Especially because the flows through the narrow channels built up by these finned rod fuels would be different from the flow characteristics in the coolant channels formed by bare rod fuels, some experiments studies to investigate the flow behaviors and structures in a finned rod bundle were done by other researchers. But because of the very complex geometries of the flow channels in the finned rod bundle which only allowed us to obtain limited information about the flow characteristics, a numerical study by a computational fluid dynamics technique has been adopted to elucidate more about such a complicated flow in a finned rod bundle. In this study, for the development of an adequate computational model to simulate such a complex geometry, a mesh sensitivity study and the effects of various turbulence models were examined. The CFD analysis results were compared with the experimental results. Some of them have a good agreement with the experimental results. All the linear eddy viscosity turbulence models could hardly predict the secondary flows near the fuel surfaces and in the sub-channel, but the RSM (Reynolds Stress Model) showed very different results from the eddy viscosity turbulence models. (author)
Additional details
Publishing Information
- Imprint Title
- NTHAS6: Proceedings of the 6th Japan-Korea symposium on nuclear thermal hydraulics and safety
- Imprint Pagination
- 818 p.
- Journal Page Range
- [6 p.]
Conference
- Title
- 6. Japan-Korea symposium on nuclear thermal hydraulics and safety
- Acronym
- NTHAS6
- Dates
- 24-27 Nov 2008
- Place
- Nago, Okinawa (Japan)
INIS
- Country of Publication
- Japan
- Country of Input or Organization
- Japan
- INIS RN
- 43053522
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- A CODES; COMPUTERIZED SIMULATION; COOLANTS; FINS; FLOW MODELS; FLUID FLOW; FUEL CHANNELS; FUEL ELEMENT CLUSTERS; FUEL RODS; HANARO REACTOR; HEAT TRANSFER; NUMERICAL ANALYSIS; THERMAL HYDRAULICS; TURBULENCE
- Descriptors DEC
- COMPUTER CODES; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; FLUID MECHANICS; FUEL ASSEMBLIES; FUEL ELEMENTS; HYDRAULICS; IRRADIATION REACTORS; ISOTOPE PRODUCTION REACTORS; MATERIALS TESTING REACTORS; MATHEMATICAL MODELS; MATHEMATICS; MECHANICS; POOL TYPE REACTORS; REACTOR CHANNELS; REACTOR COMPONENTS; REACTORS; RESEARCH AND TEST REACTORS; RESEARCH REACTORS; SIMULATION; TEST FACILITIES; TEST REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- This USB flash memory can be used for WINDOWS 2000/XP, MACINTOSH 9.x/10.x; Acrobat Reader is included; Data in PDF format, Folder Name: FullPaper, Paper ID: N6P1029.pdf; 6 refs., 11 figs., 1 tab.