Core Thermal Hydraulics Design



In the previous chapter we explored the methodology for determining the temperature field for a single fuel pin. Since a typical fast reactor core comprises several thousand fuel pins clustered in groups of several hundred pins per assembly, a complete thermal-hydraulic analysis requires knowledge of coolant distributions and pressure losses throughout the core. This chapter will address these determinations.


Pressure Drop Fuel Assembly Control Volume Approach Direct Uncertainty Inlet Plenum 
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  1. 1.
    E. H. Novendstern, “Turbulent Flow Pressure Drop Model for Fuel Rod Assemblies Utilizing a Helical Wire-Wrap Spacer System,” Nucl. Eng. Des., 22 (1972) 19–27.CrossRefGoogle Scholar
  2. 2.
    W. A. Sangster, “Calculation of Rod Bundle Pressure Loss,” Paper 68-WA/HT-35, ASME, New York, NY, 1968.Google Scholar
  3. 3.
    C. Chiu, W. M. Rohsenow, and N. E. Todreas, Flow Split Model for LMFBR Wire Wrapped Assemblies, COO-2245-56TR, Massachusetts Institute of Technology, Cambridge, April 1978.Google Scholar
  4. 4.
    J. T. Hawley, C. Chiu, W. M. Rohsenow, and N. E. Todreas, “Parameters for Laminar, Transition, and Turbulent Longitudinal Flows in Wire Wrap Spaced Hexagonal Arrays,” Topical Meeting on Nuclear Reactor Thermal Hydraulics, Saratoga, NY, 1980.Google Scholar
  5. 5.
    D. S. Rowe, COBRA-IIIC: A Digital Computer Program for Steady Slate and Transient Thermal Hydraulic Analysis of Rod Bundle Nuclear Fuei Elements, BNWL-1695, Battelle Pacific Northwest Laboratories, March 1973. See also T. L. George, K. L. Basehore, C. L. Wheeler, W. A. Prather, and R. E. Masterson, COBRA-WC: A Version of COBRA for Single-Phase Multiassembly Thermal Hydraulic Transient Analysis, Pacific Northwest Laboratory, PNL-3259, Richland, Washington, July 1980.Google Scholar
  6. 6.
    W. T. Sha, R. C. Schmitt, and P. R. Huebotter, “Boundary-Value Thermal Hydraulic Analysis of a Reactor Fuel Rod Bundle,” Nucl. Sci. Eng., 59 (1976) 140–160.Google Scholar
  7. 7.
    M. D. Carelli and C. W. Bach, “LMFBR Core Thermal Hydraulic Analysis Accounting for Interassembly Heat Transfer,” Trans. ANS, 28 (June 1978) 560–562.Google Scholar
  8. 8.
    J. L. Wantland, “ORRIBLE—A Computer Program for Flow and Temperature Distribution in 19-Rod LMFBR Fuel Subassemblies,” Nucl. Tech., 24 (1974) 168–175.Google Scholar
  9. 9.
    E. U. Khan, W. M. Rohsenow, A. A. Sonein, and N. E. Todreas, “A Porous Body Model for Predicting Temperature Distribution in Wire-Wrapped Fuel Rod Assemblies,” Nucl. Eng. Des., 35 (1975) 1–12.CrossRefGoogle Scholar
  10. 10.
    E. U. Khan, W. M. Rohsenow, A. A. Sonein, and N. E. Todreas, “A Porous Body Model for Predicting Temperature Distribution in Wire Wrapped Rod Assemblies in Combined Forced and Free Convection,” Nucl. Eng. Des., 35 (1975) 199–211.CrossRefGoogle Scholar
  11. 11.
    B. Chen and N. E. Todreas, Prediction of Coolant Temperature Field in a Breeder Reactor Including Interassembly Heat Transfer, COO-2245-20TR, Massachusetts Institute of Technology, Cambridge, MA, 1975.Google Scholar
  12. 12.
    J. N. Lillington, SABRE-3-A Computer Program for the Calculation of Steady State Boiling in Rod-Clusters, AEEW-M-1647, United Kingdom Atomic Energy Authority, 1979.Google Scholar
  13. 13.
    J. E. Meyer, Conservation Laws in One-Dimensional Hydrodynamics, WAPD-BT-20, Westinghouse Electric Corp., Bettis Atomic Power Laboratory, Pittsburgh, PA, September 1960.Google Scholar
  14. 14.
    R. E. Masterson and L. Wolf, “An Efficient Multidimensional Numerical Method for the Thermal-Hydraulic Analysis of Nuclear Reactor Cores,” Nucl. Sci. Eng., 64 (1977) 222–236.Google Scholar
  15. 15.
    J. T. Rogers and N. E. Todreas, “Coolant Interchannel Mixing in Reactor Fuel Rod Bundles Single-Phase Coolants,” Symposium on Heat Transfer in Rod Bundles, ASME, New York, NY, (1965), 1–56.Google Scholar
  16. 16.
    T. Ginsberg, “Forced-Flow Interchannel Miving Model for Fuel Rod Assemblies Utilizing a Helical Wire-Wrap Spacer System,” Nucl. Eng. Des., 22 (1972) 28–42.CrossRefGoogle Scholar
  17. 17.
    M. W. Cappiello and T. F. Cillan, Core Engineering Technical Program Progress Report, Jan-March 1977, HEDL-TME 77-46 (July 1977), Hanford Engineering Development Laboratory, Richland, WA.Google Scholar
  18. 18.
    Chaumont, Clauzon, Delpeyroux, Estavoyer, Ginier, Marmonier, Mougniot, “Conception du Coeur et des Assemblages d’une Grande Centrale a Neutrons Rapides.” Conf. Nucleaire Europeene, Paris, April 1975.Google Scholar
  19. 19.
    S. F. Wang and N. E. Todreas, Input Parameters to Codes Which Analyze LMFBR Wire-Wrapped Bundles, Rev. 1, COO-2245-17TR, Massachusetts Institute of Technology, Cambridge, MA, May 1979.Google Scholar
  20. 20.
    Preliminary Safety Analysis Report, Clinch River Breeder Reactor Plant, Project Management Corporation, 1974.Google Scholar
  21. 21.
    A. N. de Stordeur, “Drag Coefficients for Fuel-Element Spacers,” Nucleonics, 19 (1961) 74–79.Google Scholar
  22. 22.
    K. Rehme, The Measurement of Friction Factors for Axial Flow Through Rod Bundles with Different Spacers, Performed on the INR Test Rig, EURFNR-142P, November 1965.Google Scholar
  23. 23.
    T. C. Reihman, An Experimental Study of Pressure Drop in Wire Wrapped FFTF Fuel Assemblies, BNWL-1207, Richland, WA, September 1969.Google Scholar
  24. 24.
    W. Baumann, V. Casal, H. Hoffman, R. Moeller, and K. Rust, Fuel Elements with Spiral Spacers for Fast Breeder Reactors, EURFNR-571, April 1968.Google Scholar
  25. 25.
    R. A. Jaross and F. A. Smith, Reactor Development Program Progress Report, ANL-7742, Argonne National Laboratory, Argonne, IL, 1970, 30.Google Scholar
  26. 26.
    E. Bubelis and M. Schikorr, “Review and proposal for best fit of wire-wrapped fuel bundle friction factor and pressure drop predictions using various existing correlations,” Nucl. Eng. Des., 238 (2008) 3299–3320.CrossRefGoogle Scholar
  27. 27.
    G. J. Calamai, R. D. Coffield, L. J. Ossens, J. L. Kerian, J. V. Miller, E. H. Novendstern, G. H. Ursin, H. West, and P. J. Wood, Steady Stare Thermal and Hydraulic Characteristics of the FFTF Fuel Assemblies, ARD-FRT-1582, Westinghouse Electric Corp., Sunnyvale, CA, June 1974.Google Scholar
  28. 28.
    M. D. Carelli and R. A. Markley, “Preliminary Thermal-Hydraulic Design and Predicted Performance of the Clinch River Breeder Reactor Core,” Nat. Heat Transfer Conf., ASME Paper 75-HT-71, ASME, New York (1975).Google Scholar
  29. 29.
    M. D. Carelli and A. J. Friedland, “Hot channel factors for rod temperature calculations in LMFBR assemblies,” Nucl. Eng. Des., 62 (1980) 155–180.CrossRefGoogle Scholar
  30. 30.
    Y. S. Tang, R. K. Coffield, Jr., and R. A. Markley, Thermal Analysis of Liquid Metal Fast Breeder Reactors, The American Nuclear Society, La Grange Park, IL, 1978.Google Scholar
  31. 31.
    L. Walters, D. Wade, and G. Hofman, “An Innovative Particulate Metallic Fuel for Next Generation Nuclear Energy,” Proceedings of ICAPP-10, San Diego, CA, June 13–17, 2010, Paper 10356.Google Scholar

Copyright information

© Springer Science+Business Media, LLC 2012

Authors and Affiliations

  1. 1.Pacific Northwest National Laboratory (PNNL)RichlandUSA
  2. 2.NuScale PowerCorvallisUSA

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