CFD investigation of flow in the MATIS-H test facility
-
L. Tiborcz
Abstract
In this paper steps of calculations for a bare rod bundle and so-called swirl type spacer grid are showed and the effects of choice of model details are examined. The modeled pin bundle and spacer grid geometry are related to rectangular western type PWR fuel assemblies. First a CFD model has been developed for a subchannel of the MATIS-H test facility in order to carry out a mesh independence study. Based on the results a model for the half cross-section of the test bundle has been developed and calculations are carried out with different turbulence models. Translational periodicity is applied in axial direction in order to calculate a fully developed flow, which will be used as inlet boundary condition for spacer grid calculations. Results are compared with LDA (Laser Doppler Anemometry) measurements published in the frame of the OECD NEA MATIS-H benchmark. As a last step a model for rod bundle incorporating swirl type spacer grid is developed and calculations are carried out with different turbulence models using the results of the bare rod bundle simulations as inlet boundary conditions.
Kurzfassung
In diesem Beitrag werden Rechnungen zu einem Stabbündel und einem sogenannten Swirl-Abstandshalter vorgestellt. Für diese an rechteckige DWR-Brennelemente angelehnte Geometrie des Bündels werden die Einflüsse der Modellauswahl auf die Ergebnisse untersucht. Dazu wurde zuerst ein Unterkanal der MATIS-H-Versuchsanlage mit einem CFD-Modell nachgebildet und Untersuchungen zur Bestimmung einer netzunabhängigen CFD-Rechnung durchgeführt. Darauf aufbauend wurde die halbe Querschnittsfläche des Testbündels abgebildet und Rechnungen mit verschiedenen Turbulenzmodellen wurden durchgeführt. Die Ergebnisse wurden mit LDA Messungen, die im Rahmen des OECD NEA MATIS-H Benchmarks veröffentlicht wurden, verglichen. Im letzten Schritt wurde das Modell um die Abstandshalter erweitert. Mit verschiedenen Turbulenzmodellen und den Ergebnissen der ersten Rechnungen zu den Brennelementen ohne Abstandshalter als Randbedingung der Eintrittsströmung wurden weitere Rechnungen durchgeführt.
References
1 Brandt, T.; Toppila, T.: CFD study on mixing in VVER-440 fuel rod bundle with guiding vanes in spacer grids. 17th International Conference on Nuclear Engineering (ICONE-17), Brussels, Belgium, July 12–16, 200910.1115/ICONE17-75316Suche in Google Scholar
2 ANSYS Help System: CFX Theory Guide. CFX Modeling Guide 13.0, 2010Suche in Google Scholar
3 OECD/NEA: MATIS-H benchmark: Final benchmark specifications. Paris, France, Tech. Rep., July 15, 2011Suche in Google Scholar
4 Tóth, S.; Aszódi, A.: CFD analysis of flow field in a triangular rod bundle, Nuclear Engineering and Design240 (2010) 352–36310.1016/j.nucengdes.2008.08.020Suche in Google Scholar
5 Golibrodo, L. A.; Strebnev, N. A.; Kurnosov, M. M.; Galkin, I. U.; Vdovkina, I. K.: CFD simulation of turbulent flow structures in a rod bundle array with the split-type spacer grids. In: Proc. of Experimental Validation and Application of CFD and CMFD Codes in Nuclear Reactor Technology (CFD4NRS-4), Daejon, Korea, September 10–12, 2012Suche in Google Scholar
6 Lee, J. R.; Kim, J.; Song, C.-H.: Synthesis of the OECD/NEA-KAERI rod bundle CFD benchmark exercise. In: Proc. of Experimental Validation and Application of CFD and CMFD Codes in Nuclear Reactor Technology (CFD4NRS-4), Daejeon, Korea, September 10–12, 2012Suche in Google Scholar
7 Chang, S.-K.; Kim, S.; Song, C.-H.: OECD/NEA-KAERI rod bundle CFD benchmark exercise test. In: Proc. of Experimental Validation and Application of CFD and CMFD Codes in Nuclear Reactor Technology (CFD4NRS-4), Daejon, Korea, September 10–12, 2012.Suche in Google Scholar
8 Házi, G.: On turbulence models for rod bundle flow computations. Annals of Nuclear Energy32 (2005) 755–76110.1016/j.anucene.2004.12.012Suche in Google Scholar
9 Horváth, Á.; Dressel, B.: Numerical simulation of square rod arrayed rod bundles. Nuclear Engineering and Design247 (2012) 168–18210.1016/j.nucengdes.2012.03.004Suche in Google Scholar
10 Versteeg, H. K.; Malalasekera, W.: An Introduction to Computational Fluid Dynamics: The Finite Volume Method. Prentice Hall, 2007Suche in Google Scholar
11 Pope, S. B.: Turbulent Flows. Cambridge University Press, New York, 2000, p. 358–35910.1017/CBO9780511840531Suche in Google Scholar
12 Liu, C. C.; Ferng, Y. M.: Numerically simulating the thermal-hydraulic characteristics within the fuel rod bundle using CFD methodology. Nuclear Engineering and Design240 (2010) 3078–308610.1016/j.nucengdes.2010.05.021Suche in Google Scholar
© 2014, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Editorial
- Technical Contributions/Fachbeiträge
- Highly enriched alternatives of VVER-440 fuel assembly
- “FULL-CORE” VVER-440 calculation benchmark
- Development of approximation method to evaluate isotopic composition of burnt fuel
- Fuel assembly burnup calculations for VVER fuel assemblies with the MONTE CARLO code SERPENT
- Solution of the CB6 benchmark on VVER-440 final disposal using the Serpent reactor physics code
- Development and verification of new nodal methods in the KIKO3DMG code
- HPLWR fine mesh core analysis
- Assessment of reactor scram effectiveness based on measured worth of separate CR groups
- Engineering factors of the macrocode MOBY-DICK
- CFD investigation of flow in the MATIS-H test facility
- Investigation of the hot-channel calculation methodology in case of shroud-less assemblies
- Assessment of the uncertainties of COBRA sub-channel calculations by using a PWR type rod bundle and the OECD NEA UAM and the PSBT benchmarks data
- Comparison analysis of effectiveness of diagnostic methods of local coolant boiling in WWER core
- Sensitivity of hydrodynamic parameters' distributions in VVER-1000 reactor pressure vessel (RPV) with respect to uncertainty of the local hydraulic resistance coefficients
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Editorial
- Technical Contributions/Fachbeiträge
- Highly enriched alternatives of VVER-440 fuel assembly
- “FULL-CORE” VVER-440 calculation benchmark
- Development of approximation method to evaluate isotopic composition of burnt fuel
- Fuel assembly burnup calculations for VVER fuel assemblies with the MONTE CARLO code SERPENT
- Solution of the CB6 benchmark on VVER-440 final disposal using the Serpent reactor physics code
- Development and verification of new nodal methods in the KIKO3DMG code
- HPLWR fine mesh core analysis
- Assessment of reactor scram effectiveness based on measured worth of separate CR groups
- Engineering factors of the macrocode MOBY-DICK
- CFD investigation of flow in the MATIS-H test facility
- Investigation of the hot-channel calculation methodology in case of shroud-less assemblies
- Assessment of the uncertainties of COBRA sub-channel calculations by using a PWR type rod bundle and the OECD NEA UAM and the PSBT benchmarks data
- Comparison analysis of effectiveness of diagnostic methods of local coolant boiling in WWER core
- Sensitivity of hydrodynamic parameters' distributions in VVER-1000 reactor pressure vessel (RPV) with respect to uncertainty of the local hydraulic resistance coefficients
