THE APPLICABILITY OF CFD TO SIMULATE AND STUDY THE MIXING PROCESS AND THE THERMO-HYDRAULIC CONSEQUENCES OF A MAIN STEAM LINE BREAK (MSLB) IN PWR MODEL

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Ezddin HUTLI Istvan FARKAS Tatiana FARKAS Takács ANTAL Ivan TOTH Guba ATTILA6

Abstract

This paper focuses on the validation and applicability of Computational Fluid Dynamics (CFD) to simulate and analyze the thermo-hydraulic consequences of a Main Steam Line Break (MSLB). Extensive validation  data come from experiments performed using the Rossendorf Coolant  Mixing Model (ROCOM) facility. For the calculation, the range of 9 to 12 million hexahedral cells was constructed to capture all details in the interrogation domain in the system. The analysis was performed by running a time-dependent calculation, Detailed analyses were made at different cross-sections in the system to evaluate not only the value of the maximum and minimum temperature, but also the location and the time at which it occurs during the transient which is considered to be indicator for  the quality of mixing in the system. CFD and experimental results were qualitatively compared; mixing in the cold legs with Emergency Core Cooling Systems (ECCS) was overestimated. This could be explained by the sensitivity to the boundary conditions. In the downcomer, the experiments displayed higher mixing: by our assumption this related to the dense measurement grid (they were not modelled). The temperature distribution in the core inlet plane agreed with the measurement results. Minor deviations were seen in the quantitative comparisons: the maximum temperature difference was 2ºC.

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How to Cite
HUTLI, Ezddin et al. THE APPLICABILITY OF CFD TO SIMULATE AND STUDY THE MIXING PROCESS AND THE THERMO-HYDRAULIC CONSEQUENCES OF A MAIN STEAM LINE BREAK (MSLB) IN PWR MODEL. Thermal Science, [S.l.], mar. 2017. ISSN 2334-7163. Available at: <http://thermal-science.tech/journal/index.php/thsci/article/view/2236>. Date accessed: 17 aug. 2017. doi: https://doi.org/10.2298/TSCI151126251F.
Section
Articles
Received 2017-03-06
Accepted 2017-03-13
Published 2017-03-13

References

[1] Rohde, U., et al., 2007. Fluid Mixing and Flow Distribution in a Primary Circuit of a Nuclear Pressurized Water Reactor-Validation of CFD Codes. Nuclear Engineering and Design. 237, 1639–1655.
[2] Helmholtz-Zentrum Dresden-Rossendor, Experimental studies of the fluid dynamics during a pressurized thermal shock (PTS) https://www.hzdr.de/db/Cms?pOid=25292&pNid=3016
[3] Farkas, T., Tóth I., 2010. Fluent Analysis of a ROSA Cold Leg Stratification Test. Nuclear Engineering and Design. 240, 2169–2175.
[4] Kiger, K.T., Gavelli F., 2001. Boron Mixing in Complex Geometries: Flow Structure Details. Nuclear Engineering and Design. 208, 67–85.
[5] Rohde, U., et al., 2005. Fluid Mixing and Flow Distribution in the Reactor Circuit, Measurement Data Base. Nuclear Engineering and Design. 235, 421– 443. http://www.enea.it/it/Ricerca_sviluppo/documenti/ricerca-di-sistema-elettrico/nuovo-nucleare- fissione/lp2/rds-109-lp2.pdf
[6] IAEA-TECDOC-1627, 2010. Pressurized Thermal Shock in Nuclear Power Plants: Good Practices for Assessment, Deterministic Evaluation for the Integrity of Reactor Pressure Vessel. International Atomic Energy Agency Vienna, http://www-pub.iaea.org/MTCD/publications/PDF/te_1627_web.pdf
[7] Jiejin, C., Tadashi, W., 2011. Numerical Simulation of Thermal Stratification in Cold Legs by Using OpenFOAM . Progress in Nuclear Science and Technology. 2, 107–113.
[8] Nobuchika, K., et al., 2006. Spectra Thermal Fatigue Tests under Frequency Controlled Fluid Temperature Variation: Transient Temperature Measurement Tests, ASME Pressure Vessels and Piping/ICPVT-11 Conference, 857–864.
[9] Courtin, S., 2013. High Cycle Thermal Fatigue Damage Prediction in Mixing Zones of Nuclear Power Plants: Engineering Issues Illustrated on the FATHER Case, 5th Fatigue Design Conference, Fatigue Design, Procedia Engineering. 66, 240–249.
[10] Gottlasz, V., et al., 2013. Measuring of velocity and temperature field in a model of reactor vessel downcomer and cold-legs inlets, 10th International Symposium on Particle Image Velocimetry- PIV13Delft, The Netherlands. http://www.google.ch/url?sa=t&rct=j&q=&esrc=s&source=web&cd=3&ved=0CDMQFjAC&url= ht%3A%2F%2Frepository.tudelft.nl%2Fassets%2Fuuid%3A6ea59a1e-c4f8-48b0-b5f6 c250b6b4e09a%2FA025_paper.pdf&ei=taZXUWNI4HnygPErgI&usg=AFQjCNEyhfrNWA5b67S RsSROlQO8ZszdCw
[11] Hutli, E., et al., 2013. Investigation of Mixing Flow Process Using PIV and PLIF Techniques. Advanced Materials Research. 816 – 817, 1054 –1058.
[12] Pochet, G., et al., 2011. CFD Simulations of Buoyancy Driven Flow Mixing Experiments Performed at the Rocom Facility, the14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14). 1–12.
[13] Yamaji, B., Aszodi, A., 2006. CFD Analysis of Coolant Mixing in VVER-1000 Pressure Vessel. Proceedings of the 16th Symposium of AER on VVER Reactor Physics and Reactor Safety, Bratislava, Slovakia. 483 – 491.
[14] Hohne, T., Sören, K., 2007. Modeling of a Buoyancy-Driven Flow Experiment in Pressurized Water Reactors Using CFD Methods. Nuclear Engineering and Technology. 39, 327–336.
[15] Vaibar, R., Höhne, T., 2009. Buoyancy Driven Flow in Reactor Safety. Applied and Computational Mechanics.3, 22–232.
[16] Gango, P., 1997. Numerical Boron Mixing Studies for Loviisa Nuclear Power Plant. Nuclear Engineering Design. 177, 239–254.
[17] Hohne, T., et al., 2008. Boron Dilution Transients During Natural Circulation Flow in PWR Experiments and CFD Simulations. Nuclear Engineering Design. 238, 1987–1995.
[18] Dury, T.V., et al., 2008. CFD Simulation of the Vattenfall 1/5th-Scalepwr Model for Boron Dilution. Nuclear Engineering Design. 238, 577–589.
[19] Hohne, T., et al., 2006. Modeling of a Buoyancy-Driven Flow Experimental the ROCOM Test Facility Using the CFD Codes CFX-5 and Trio-U. Nuclear Engineering Design. 236, 1309 – 1325.
[20] Loginov, M., Komen, E., Kuczaj, A., 2010. Application of large-Eddy Simulation to Pressurized Thermal Shock: A grid Resolution Study. Nuclear Engineering Design. 240, 2034 – 2045.
[21] Loginov, M., et al., 2011. Application of Large-Eddy Simulation to pressurized Thermal Shock: Assessment of the accuracy. Nuclear Engineering Design. 241, 3097–3110.
[22] Ducros, F., et al., 2010. Verification and Validation Considerations Regarding the Qualification of Numerical Schemes for LES for Dilution Problems. Nuclear Engineering Design. 240, 2123 –2130
[23] Jayaraju, S.T, et al., 2013. Large Eddy Simulation for an Inherent Boron D ilution Transient. Nuclear Engineering and Design. 262, 484 – 498.
[24] Hutli, E., et al., 2015. Investigation of mixing coolant in a model of reactor vessel down-comer and in cold leg inlets using PIV, LIF, and CFD techniques, Thermal Science,doi:10.2298/ TSCI14091 5121H. http://www.doiserbia.nb.rs/img/doi/0354-9836/2015%20OnLine-First/0354- 98361500121H.pdf

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