3.17. References

  1. M. Lees, “An Extrapolated Crank-Nicholson Difference Scheme for Quasilinear Parabolic Equations” Proc. Symp. On Nonlinear Partial Differential Equations, W. F. Ames, Ed., Academic Press, 1966.

  2. F. E. Dunn, et al., “The SAS2A LMFBR Accident Analysis Computer Code,” ANL-8138, Argonne National Laboratory, 1974.

  3. “A Preliminary User’s Guide to Version 1.0 of the SAS3D LMFBR Accident Analysis Computer Code,” J. E. Cahalan and D. R. Ferguson, Eds.; Available through the Reactor Analysis Division, Argonne National Laboratory, 1977.

  4. W. M. Rohsenow and H. Y. Choi, Heat, Mass and Momentum Transfer, p. 189, Prentice-Hall, 1961.

  5. W. M. Rohsenow and J. P. Hartnett, Eds., Handbook of Heat Transfer, p. 7-33, McGraw-Hill Book company, New York, 1973.

  6. William T. Sha, Frank J. Goldner, Paull R. Heubotter, and Robert C. Schmitt, “Thermal Hydraulic Multichannel Analysis,” Proceedings of the International Meeting on Reactor Heat Transfer, Karlsruhe, Germany, p. 180, October 9-11, 1973.

  7. D. Mohr, L. Chang, and H. P. Planchon, “Validation of the HOTCHAN Code for Analyzing the EBR-II Core Following an Unprotected Loss of Flow,” Trans. Am. Nucl. Soc., vol. 57, p. 318, 1988.

  8. Y. M. Kwon, Y. B. Lee, W. P. Chang and D. Hahn, “SSC-K Code User’s Manual (Rev. 1)”, KAERI/TR-2014, KAERI, Daejeon, Korea, 2002.

  9. C. W. Stewart, C. L. Wheeler, R. J. Cena, C.A. McMonagle, J.M. Cuta and D.S. Trent, “COBRA-IV: the Model and the Method,” BNWL-2214, Battelle Pacific Northwest Laboratories, Richland, Washington, 1977.

  10. K. L. Basehore and N. E. Todreas, “SUPERENERGY-2: a MULTIASSEMBLY Steady-State Computer Code for lmfbr core Thermal-hydraulics Analysis,” pnl-3379, pacific northwest laboratory, Hanford, Washington, 1980.

  11. P. M. Magee, A. E. Dubberley, A. J. Lipps and T. Wu, “Safety Performance of the Advanced Liquid Metal Reactor,” Proceedings of ARS ‘94 International Topical Meeting on Advanced Reactors Safety, Pittsburg, Pennsylvania, April 17-21, vol. 2, pp. 826-833, 1994.

  12. S. K. CHENG and N. E. Todreas, “Hydrodynamic Models and Correlations for Bare and Wire-Wrapped Hexagonal Rod Bundles - Bundle Friction Factors, Subchannel Friction Factors and Mixing Parameters,” Nucl. Eng. And Design, vol. 92, p. 227-251, 1986.

  13. “Nuclear Systems Materials Handbook”, TID-26666, Hanford Engineering Development Laboratory.

  14. S. Y. Ogawa, E. A. Lees and M. F. Lyons, “Power Reactor High Performance UO2 Program, Fuel Design Summary and Progress Status,” GEAP-5591, General Electric Company, 1968.

  15. P. Verbeek and H. Hoppe, “COMETHE-IIIJ: A Computer Code for Predicting Mechanical and Thermal Behavior of a Fuel Pin, Part 1: General Description,” INBFR-107 (original report BN-7609-07), Belgonucleaire, Brussels, Belgium, 1977.

  16. S.K. Chen, Y.M. Chen, and N.E. Todreas, “The upgraded Cheng and Todreas correlation for pressure drop in hexagonal wire-wrapped rod bundles,” Nucl. Eng. And Design, vol. 335, p. 356-373, 2018.