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Heat and Mass Transfer

, Volume 53, Issue 2, pp 537–551 | Cite as

Natural and mixed convection in the cylindrical pool of TRIGA reactor

  • R. Henry
  • I. Tiselj
  • M. Matkovič
Original
  • 211 Downloads

Abstract

Temperature fields within the pool of the JSI TRIGA MARK II nuclear research reactor were measured to collect data for validation of the thermal hydraulics computational model of the reactor tank. In this context temperature of the coolant was measured simultaneously at sixty different positions within the pool during steady state operation and two transients. The obtained data revealed local peculiarities of the cooling water dynamics inside the pool and were used to estimate the coolant bulk velocity above the reactor core. Mixed natural and forced convection in the pool were simulated with a Computational Fluid Dynamics code. A relatively simple CFD model based on Unsteady RANS turbulence model was found to be sufficient for accurate prediction of the temperature fields in the pool during the reactor operation. Our results show that the simple geometry of the TRIGA pool reactor makes it a suitable candidate for a simple natural circulation benchmark in cylindrical geometry.

Keywords

Computational Fluid Dynamic Natural Convection Computational Fluid Dynamic Simulation Computational Fluid Dynamic Model Reactor Core 
These keywords were added by machine and not by the authors. This process is experimental and the keywords may be updated as the learning algorithm improves.

List of symbols

α

Thermal diffusivity (m2 s−1)

β

Coefficient of thermal expansion (K−1)

ε

Turbulent dissipation (J kg−1 s−1)

λ

Thermal conductivity (W m−1 K−1)

ρ

Density (kg m−3)

τ

Time constant of the thermocouple (s)

ν

Kinematic viscosity (m2 s−1)

δ

Error (δx—error in x coordinate)

Δ

Change, difference (Δt—change in time t)

cp

Specific heat capacity (J kg−1 K−1)

d

Diameter of the thermocouple (m)

g

Gravity acceleration (m s−2)

h

Heat transfer coefficient (W m−2 K−1)

k

Turbulent kinetic energy (J)

l

Fuel element half-length (m)

m

Mass (kg)

\(\dot{m}\)

Mass flow rate (kg s−1)

q″

Heat flux (W m−2)

t

Time (s)

v

Velocity (m/s)

x, y, z

Cartesian coordinate components (m)

xi

i-th spatial coordinate (m)

C

Heat capacity of the pool (J K−1)

Lc

Characteristic length (m)

Ra

Rayleigh number

S

Surface of the thermocouple (m2)

T

Temperature (°C)

Text

Temperature measured by the thermocouple (°C)

Vpool

Volume of the pool (m3)

\(\dot{V}\)

Volumetric flow rate (m3 s−1)

Notes

Acknowledgments

The authors gratefully acknowledge the valuable contributions of the reactor operator staff at JSI, in particular from Anže Jazbec, Darko Kavšek, Sebastjan Rupnik and Marko Rosman, for their help from the conception to the realization of the experiment. Research was founded by P2-0026 research program Nuclear engineering.

References

  1. 1.
    Papanicolaou E, Belessiotis V (2002) Transient natural convection in a cylindrical enclosure at high Rayleigh numbers. Int J Heat Mass Transf 45(7):1425–1444. doi: 10.1016/s0017-9310(01)00258-7 CrossRefzbMATHGoogle Scholar
  2. 2.
    Oliveski RDC, Krenzinger A, Vielmo HA (2003) Cooling of cylindrical vertical tanks submitted to natural internal convection. Int J Heat Mass Transf 46(11):2015–2026. doi: 10.1016/s0017-9310(02)00508-2 CrossRefGoogle Scholar
  3. 3.
    Hmouda I, Rodriguez I, Bouden C, Oliva A (2010) Unsteady natural convection cooling of a water storage tank with an internal gas flue. Int J Therm Sci 49(1):36–47. doi: 10.1016/j.ijthermalsci.2009.05.011 CrossRefGoogle Scholar
  4. 4.
    Fan J, Furbo S (2012) Buoyancy driven flow in a hot water tank due to standby heat loss. Sol Energy 86(11):3438–3449. doi: 10.1016/j.solener.2012.07.024 CrossRefGoogle Scholar
  5. 5.
    Agency IAEA (2005) Natural circulation in water cooled nuclear power plants. IAEA-TECDOC-1474Google Scholar
  6. 6.
    Todreas NE, Kazimi MS (2015) Nuclear systems, volume I: thermal hydraulic fundamentals, Revised 2nd edn. CRC PressGoogle Scholar
  7. 7.
    Todreas NE, Kazimi MS (2001) Nuclear systems, volume II: elements of thermal hydraulic design. Taylor & Francis, Park DriveGoogle Scholar
  8. 8.
    Prošek A, Volkanovski A (2015) RELAP5/MOD3.3 analyses for prevention strategy of extended station blackout. J Nucl Eng Radiat Sci 1(4):041016. doi: 10.1115/1.4030834 CrossRefGoogle Scholar
  9. 9.
    Hung T-C, Dhir VK, Pei B-S, Chen Y-S, Tsai FP (2013) The development of a three-dimensional transient CFD model for predicting cooling ability of spent fuel pools. Appl Therm Eng 50(1):496–504. doi: 10.1016/j.applthermaleng.2012.06.042 CrossRefGoogle Scholar
  10. 10.
    Agency IAEA (2009) Passive safety systems and natural circulation in water cooled nuclear power plants. IAEA-TECDOC-1624Google Scholar
  11. 11.
    Krepper E, Beyer M (2010) Experimental and numerical investigations of natural circulation phenomena in passive safety systems for decay heat removal in large pools. Nucl Eng Des 240(10):3170–3177. doi: 10.1016/j.nucengdes.2010.05.050 CrossRefGoogle Scholar
  12. 12.
    Tenchine D, Pialla D, Fanning TH, Thomas JW, Chellapandi P, Shvetsov Y, Maas L, Jeong HY, Mikityuk K, Chenu A, Mochizuki H, Monti S (2013) International benchmark on the natural convection test in Phenix reactor. Nucl Eng Des 258:189–198. doi: 10.1016/j.nucengdes.2013.02.010 CrossRefGoogle Scholar
  13. 13.
    de Kruijf WJM, Ketelaar KCJ, Avakian G, Gubernatis P, Caruge D, Manera A, Van der Hagen THJJ, Yadigaroglu G, Dominicus G, Rohde U, Prasser HM, Castrillo F, Huggenberger M, Hennig D, Munoz-Cobo JL, Aguirre C (2003) Planned experimental studies on natural-circulation and stability performance of boiling water reactors in four experimental facilities and first results (NACUSP). Nucl Eng Des 221(1–3):241–250. doi: 10.1016/s0029-5493(02)00338-2 CrossRefGoogle Scholar
  14. 14.
    Espinosa-Paredes G, Verma SP, Vazquez-Rodriguez A, Nunez-Carrera A (2010) Mass flow rate sensitivity and uncertainty analysis in natural circulation boiling water reactor core from Monte Carlo simulations. Nucl Eng Des 240(5):1050–1062. doi: 10.1016/j.nucengdes.2010.01.012 CrossRefGoogle Scholar
  15. 15.
    Snoj L, Trkov A, Jacimovic R, Rogan P, Zerovnik G, Ravnik M (2011) Analysis of neutron flux distribution for the validation of computational methods for the optimization of research reactor utilization. Appl Radiat Isot Incl Data Instrum Methods Use Agric Ind Med 69(1):136–141. doi: 10.1016/j.apradiso.2010.08.019 Google Scholar
  16. 16.
    Ravnik M, Jeraj R (2003) Research reactor benchmarks. Nucl Sci Eng 145(1):145–152Google Scholar
  17. 17.
    Umar E, Kamajaya K, Tandian NP, Hardianto T, Suwono A (2006) An experimental study of natural convection in the hottest channel of TRIGA 2000 kW. In: Association AN (ed) Pacific basin nuclear conference SydneyGoogle Scholar
  18. 18.
    Henry R, Tiselj I, Snoj L (2015) Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP. Appl Radiat Isot Incl Data Instrum Methods Use Agric Ind Med 97:140–148. doi: 10.1016/j.apradiso.2014.12.017 Google Scholar
  19. 19.
    Snoj L, Ravnik M (2008) Power peakings in mixed TRIGA cores. Nucl Eng Des 238(9):2473–2479. doi: 10.1016/j.nucengdes.2008.02.005 CrossRefGoogle Scholar
  20. 20.
    Jensen RT, Newel DL (1998) Thermal Hydraulic calculations to support to support increase in operating power in McClellen nuclear radiation center TRIGA reactor. In: International user’s seminar, College Station, TexasGoogle Scholar
  21. 21.
    Marcum WR, Woods BG, Hartman MR, Reese SR, Palmer TS, Keller ST (2009) Steady-state thermal-hydraulic analysis of the Oregon State University TRIGA reactor using RELAP5-3D. Nucl Sci Eng 162(3):261–274. doi: 10.13182/nse08-63 CrossRefGoogle Scholar
  22. 22.
    Reis PAL, Costa AL, Pereira C, Veloso MAF, Mesquita AZ, Soares HV, Barros GdP (2010) Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor. Ann Nucl Energy 37(10):1341–1350. doi: 10.1016/j.anucene.2010.05.013 CrossRefGoogle Scholar
  23. 23.
    Feltus MA, Miller WS (2000) Three-dimensional coupled kinetics/thermal-hydraulic benchmark TRIGA experiments. Ann Nucl Energy 27(9):771–790. doi: 10.1016/S0306-4549(99)00087-0 CrossRefGoogle Scholar
  24. 24.
    Huda MQ, Rahman M (2004) Thermo-hydrodynamic design and safety parameter studies of the TRIGA MARK II research reactor. Ann Nucl Energy 31(10):1101–1118. doi: 10.1016/j.anucene.2004.02.001 CrossRefGoogle Scholar
  25. 25.
    Huda MQ, Bhuiyan SI (2006) Investigation of thermohydraulic parameters during natural convection cooling of TRIGA reactor. Ann Nucl Energy 33(13):1079–1086. doi: 10.1016/j.anucene.2006.08.001 CrossRefGoogle Scholar
  26. 26.
    Saha P, Aksan N, Andersen J, Yan J, Simoneau JP, Leung L, Bertrand F, Aoto K, Kamide H (2013) Issues and future direction of thermal-hydraulics research and development in nuclear power reactors. Nucl Eng Des 264:3–23. doi: 10.1016/j.nucengdes.2012.07.023 CrossRefGoogle Scholar
  27. 27.
    Ferziger JH, Perić M (2002) Computational Methods for Fluid Dynamics. Springer, Berlin. doi: 10.1007/978-3-642-56026-2 CrossRefzbMATHGoogle Scholar
  28. 28.
    Chen SR, Lin WC, Ferng YM, Chieng CC, Pei BS (2014) Development of 3-D CFD methodology to investigate the transient thermal-hydraulic characteristics of coolant in a spent fuel pool. Nucl Eng Des 275:272–280. doi: 10.1016/j.nucengdes.2014.05.020 CrossRefGoogle Scholar
  29. 29.
    da Silva MJ, Thiele S, Höhne T, Vaibar R, Hampel U (2010) Experimental studies and CFD calculations for buoyancy driven mixing phenomena. Nucl Eng Des 240(9):2185–2193. doi: 10.1016/j.nucengdes.2009.11.023 CrossRefGoogle Scholar
  30. 30.
    Minocha N, Joshi JB, Nayak AK, Vijayan PK (2015) Numerical investigation of three-dimensional natural circulation phenomenon in passive safety systems for decay heat removal in large pools. Int J Heat Mass Transf 81:659–680. doi: 10.1016/j.ijheatmasstransfer.2014.10.007 CrossRefGoogle Scholar
  31. 31.
    Henry R, Tiselj I (2014) A new coupling CFD/Monte Carlo neutron transport scheme, application to a single fuel rod problem. In: Slovenia LNSo (ed) 23rd International conference nuclear energy for New Europe—NENE 2014, Portorož, Slovenia. Nuclear Society of Slovenia, LjubljanaGoogle Scholar
  32. 32.
    Corp. NI (2015). http://www.ni.com/pxi/
  33. 33.
    Corp. NI (2003) LabVIEW™ user manualGoogle Scholar
  34. 34.
    Inc A (2011) User’s guide, ANSYS 14.0 documentationGoogle Scholar
  35. 35.
    Landau LD, Lifshitz EM (1959) Fluid mechanics (volume 6 of a course of theoretical physics). Pergamon PressGoogle Scholar
  36. 36.
    Launder BE, Spalding DB (1974) The numerical computation of turbulent flows. Comput Methods Appl Mech Eng 3(2):269–289. doi: 10.1016/0045-7825(74)90029-2 CrossRefzbMATHGoogle Scholar
  37. 37.
    Radulovic V, Stancar Z, Snoj L, Trkov A (2014) Validation of absolute axial neutron flux distribution calculations with MCNP with 197 Au (n, gamma) 198 Au reaction rate distribution measurements at the JSI TRIGA Mark II reactor. Appl Radiat Isot Incl Data Instrum Methods Agric Ind Med 84:57–65. doi: 10.1016/j.apradiso.2013.11.027 Google Scholar
  38. 38.
    Argyropoulos CD, Markatos NC (2015) Recent advances on the numerical modelling of turbulent flows. Appl Math Model 39(2):693–732. doi: 10.1016/j.apm.2014.07.001 MathSciNetCrossRefGoogle Scholar
  39. 39.
    Henry R, Tiselj I (2013) Computational fluid dynamic model of TRIGA Mark II reactor. In: 22nd International conference nuclear energy for New Europe—NENE 2013, Bled—Slovenia. Nuclear Society of Slovenia, LjubljanaGoogle Scholar
  40. 40.
    Štancar Ž, Snoj L (2014) Thermal power calibration of the TRIGA Mark II reactor. In: Slovenia LNSo (ed) 23rd International conference nuclear energy for New Europe—NENE 2014, Portorož, Slovenia. Nuclear Society of Slovenia, LjubljanaGoogle Scholar

Copyright information

© Springer-Verlag Berlin Heidelberg 2016

Authors and Affiliations

  1. 1.Reactor Engineering Division R4Jožef Stefan InstituteLjubljanaSlovenia

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