Thermal – hydraulic analysis of fuel block in high temepera-ture reactor

• Autorzy: Królikowski I. , Kępisty G. , Orliński M.
• Języki publikacji: EN

Abstrakty

EN

Thermal hydraulic analysis of the reactor core is important since it allows to optimize the nuclear reactor operation and to avoid too high temperature in the fuel. Enhancement of the reactor core increases the safety and the efficiency of the reactor operation and it has positive impact on the logistic in the nuclear sector. The thermal analysis of the fuel block column of the high temperature reactor is presented. The 3D power density profile has been used in the thermal calculations to obtain the temperature field within the column of the nine fuel blocks. The hot spot for the critical power profile is found. Temperature profiles obtained in the analysis have been compared with the reference data to check the numerical model, which has been used in the CFD calculations. Obtained temperatures are consistent with the reference data, it proves that the numerical model is correct.

Słowa kluczowe

EN

CFD; critical temperature; high-temperature reactor; hot spot; HTR; power profile; thermal hydraulics

PL

CFD; temperatura krytyczna; reaktor wysokotemperaturowy; hot spot; HTR; profil mocy; hydraulika cieplna

• Wydawca: Sieć Badawcza Łukasiewicz - Poznański Instytut Technologiczny
• Czasopismo: Logistyka
• Rocznik: 2015
• Tom: nr 4
• Strony: 9278--9285, CD3
• Opis fizyczny: Bibliogr. 13 poz., wykr., rys.

Twórcy

autor

Królikowski I.

• AGH University of Science and Technology, Faculty of Energy and Fuels, Department of Nuclear Engineering, Kraków, Poland

autor

Kępisty G.

• AGH University of Science and Technology, Faculty of Energy and Fuels, Department of Nuclear Engineering, Kraków, Poland

autor

Orliński M.

• AGH University of Science and Technology, Faculty of Energy and Fuels, Department of Nuclear Engineering, Kraków, Poland

Bibliografia

• [1] Yan Y., Rizwan-uddin, Kim K.: A coupled CFD-system code development and application. PHYSOR, Interlaken, Switzerland, Sep. 14–19, 2008.
• [2] Reiss T., Fehér S., Czifrus S.: Coupled neutronics and thermo hydraulics calculations with burn-up for HPLWRs. Progr. Nucl. Energy, vol. 50, pp. 52–61, 2008.
• [3] Seker V., Thomas J. W., Downar T. J.: Reactor physics simulations with coupled Monte Carlo calculations and computational fluid dynamics. Proc. Int. Conf. Emerging Nuclear Energy Systems (ICENES), 2007.
• [4] Breitkreutz H., Rohrmoser A., Petry W.: 3-Dimensional coupled neutronic and thermal-hydraulic calculations for a compact core combining MCNPX and CFX. IEEE Transactions on Neclear Sciecne, Vol. 57, NO. 6, 2010.
• [5] Jianwei Hu, Rizwan-uddin.: Coupled neutronics and thermal-hydraulics simulations using MCNP and FLUENT. Advancements in Multi-Physics Reactor Simulation, USA.
• [6] Królikowski I., Cetnar J.: Neutronics and thermal hydraulics coupling for 3D reactor core modeling combining MCB and FLUENT. NUTECH, Warsaw, 2014
• [7] ANSYS Inc., DesignModeler User Guide, 2012
• [8] ANSYS Inc., ANSYS Modeling and Meshing Guide, 2005
• [9] ANSYS Inc., ANSYS FLUENT User’s Guide, 2011
• [10] J. Cetnar. (2006). User Manual for MCB 5.
• [11] MCNP – A GENERAL MONTE CARLO CODE N-PARTICLE TRANSPORT CODE, Version 5, X-5 Monte Carlo Team, 2008
• [12] Venneri Francesco et al.: High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis, Idaho National Laboratory, INL/EXT-10-19973, September 2010.
• [13] ACK Cyfronet AGH, supercomputer MARS, IBM BladeCenter HS21, grant number MNiSW/IBM_BC_HS21/AGH/064/2011.