experimental heat transfer analysis of the ipr-r1 triga reactor - CDTN

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The graph of heat transfer coefficient through the gap is shown in Fig. 4, as a function of thereactor power. This figure also shows three theoretical values recommended by GeneralAtomic for the heat transfer coefficient [2].Figure 4. Heat transfer coefficient through the gap as a function of the power and fuelelement configuration4. FUEL ROD TEMPERATURE PROFILEFrom the temperature in the center of the fuel and using the equations of conduction for thefuel element geometry, it is possible to obtain the radial temperature distribution in the fuelelement. Figure 5 shows the experimental radial profile of maximum fuel temperature inposition B1 and it is compared with the PANTERA code results [15]. The instrumented fuelelement was used to measure the fuel temperature at several reactor powers. The results arealso shown in Fig. 5.Figure 5. Experimental fuel rod radial temperature profile in position B1 at 265 kWand at other reactor powers5. CONCLUSIONSubcooled pool boiling occurs above approximately 60 kW on the cladding surface in thecentral channels of the IPR-R1 TRIGA core. However, the high heat transfer coefficient dueto subcooled boiling causes the cladding temperature be quite uniform along most of theactive fuel rod region and do not increase very much with the reactor power. The IPR-R13 rd WORLD TRIGA USERS CONFERENCE – Belo Horizonte, MG, Brazil
TRIGA Reactor normally operates in the range from 100 kW until a maximum of 250 kW.On these power levels the heat transfer regime between the clad surface and the coolant issubcooled nucleate boiling in the hottest fuel element. Boiling heat transfer is usually themost efficient heat transfer pattern in nuclear reactors core [6]. Another important aspect ofthe reactor operation safety is that it is far from the occurrence of critical heat flux [16].ACKNOWLEDGMENTSThe authors thank to the operation staff of the IPR-R1 TRIGA Reactor for their help duringthe experiments.REFERENCES1. Todreas N.E. and Kazimi M.S., Nuclear Systems I: Thermal Hydraulic Fundaments,Hemisphere Publishing Corporation, New York, (1990).2. General Atomic, Safeguards Summary Report for the New York University TRIGA MarkI Reactor. (GA-9864). San Diego, (1970).3 Dalle H.M., Neutronic Calculations of the IPR-R1 TRIGA Reactor with WIMSD4 eCITATION. M. Sc dissertation, Universidade Federal de Minas Gerais, Belo Horizonte,(in Portuguese), (1999).4 Dalle H.M., Neutronic Analyses of the IPR-R1 TRIGA Reactor with 63 Fuel ElementsConfiguration and Regulating Control Rod in Position F16, CDTN/CNEN, NI–EC3-01/03, Belo Horizonte, (in Portuguese), (2003).5 Lamarsh J.R. and Baratta A.J., Introduction to Nuclear Engineering, 3° ed., UpperSaddler River: Prendice Hall, (2001).6 Duderstadt J.J and Hamilton L.J.; Nuclear Reactor Analysis, John Wiley & Sons, Inc.New York, (1976).7. Wagner W. and Kruse A., Properties of Water and Steam – The Industrial StandardIAPWS-IF97 for the Thermodynamics Properties, Springer, Berlin, (1998).8. Tong L.S. and Weisman J., Thermal Analysis of Pressurized Water Reactors, ThirdEdition, American Nuclear Society. Illinois, (1996)9. Mesquita A. Z., Experimental Investigation on Temperatures Distributions in a ResearchNuclear Reactor TRIGA IPR-R1, Ph.D thesis, Universidade Estadual de Campinas, SãoPaulo, (in Portuguese), (2005).10. Glasstone S. and Sesonske A., Nuclear Reactor Engineering, 4 ed., Chapman and Hall,New York, NY, (1994).11. Kreith F. and Bohn M. S., Principles of Heat Transfer, 6 th ed., Brooks/Cole, New York,(2001).12. Tong L. S. and Tang Y.S, Boiling Heat Transfer and Two-Phase Flow, 2 nd . Ed. Taylor &Francis, Washington, (1997).13. Simnad M.T., Foushee F.C. and West G.B., Fuel Elements for Pulsed TRIGA ResearchReactors, Nuclear Technology, 28:31-56. (1976).14. ASME, ASME Boiler and Pressure Vessel Code, Section II − Materials, Part D,Properties”, The American Society of Mechanical Engineers, New York, (1992).15. Veloso M.A., Thermal–hydraulic Analyses of the IPR-R1 TRIGA Reactor on 250 kW,CDTN/CNEN, NI-EC3-05/05, Belo Horizonte, (in Portuguese), (2005).16. Mesquita A.Z. and Rezende H.C., Experimental Prediction of the Critical Heat Flux onthe IPR-R1 TRIGA Nuclear Reactor. Proceeding of 3 nd World Triga Users Conference,Belo Horizonte, August 22 to 25, (2006).3 rd WORLD TRIGA USERS CONFERENCE – Belo Horizonte, MG, Brazil
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Page 3 and 4: Single-Phase Region2. HEAT TRANSFER
Page 5: Figure 3. Overall fuel element ther

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