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FAST REACTOR PHYSICS AND COMPUTATIONAL METHODS

  • Yang, W.S. (Purdue University School of Nuclear Engineering)
  • Received : 2011.03.23
  • Accepted : 2011.12.12
  • Published : 2012.03.25

Abstract

This paper reviews the fast reactor physics and computational methods. The basic reactor physics specific to fast spectrum reactors are briefly reviewed, focused on fissile material breeding and actinide burning. Design implications and reactivity feedback characteristics are compared between breeder and burner reactors. Some discussions are given to the distinct nuclear characteristics of fast reactors that make the assumptions employed in traditional LWR analysis methods not applicable. Reactor physics analysis codes used for the modeling of fast reactor designs in the U.S. are reviewed. This review covers cross-section generation capabilities, whole-core deterministic (diffusion and transport) and Monte Carlo calculation tools, depletion and fuel cycle analysis codes, perturbation theory codes for reactivity coefficient calculation and cross section sensitivity analysis, and uncertainty analysis codes.

Keywords

References

  1. J. Bouchard and R. Bennett, "A New Generation of Nuclear to Lead the Way," Energy Focus, Spring 2009 Edition, 85 (2009).
  2. J. Bouchard and R. Bennett, "Generation IV Advanced Nuclear Energy Systems," Nuclear Plant Journal, 26, 2 (2008).
  3. Leonard J. Koch, EBR-II Experimental Breeder Reactor II, Argonne National Laboratory (2004).
  4. M. E. Bunker, "Early Reactors from Fermi's Water Boiler to Novel Power Prototype," Los Alamos Science, Winter/Spring 1983 Edition, 124 (1983).
  5. J. H. Kittel et al., "History of Fast Reactor Fuel Development," J. Nuclear Materials, 204, 1 (1993). https://doi.org/10.1016/0022-3115(93)90193-3
  6. C. E. Stevenson, The EBR-II Fuel Cycle Story, American Nuclear Society (1987).
  7. J. F. Sauvage, Phenix 30 Years of History: The Heart of a Reactor, CEA, France (2004).
  8. S. H. Fistedis, ed., "The Experimental Breeder Reactor-II Inherent Safety Demonstration," Nucl. Eng. Des. (Special issue), 101 (1987).
  9. "Fast Neutron Reactors (updated August 2011)," World Nuclear Association Website, http://www.world-nuclear/org/info/inf98.html.
  10. M. B. Chadwick et al., "ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology," Nucl. Data Sheets, 107, 2931 (2006). https://doi.org/10.1016/j.nds.2006.11.001
  11. C. E. Till, et al., "Fast Breeder Reactor Studies," ANL-80-40, Argonne National Laboratory (1980).
  12. A. E. Waltar and A. E. Reynolds, Fast Breeder Reactors, Pergamon Press, New York, New York (1981).
  13. Fast Reactor Database, IAEA-TECDOC-866, International Atomic Energy Agency (1996).
  14. Y. I. Chang, et al., "Advanced Burner Test Reactor Preconceptual Design Report," ANL-ABR-1 (ANL-AFCI 173), Argonne National Laboratory, 2006.
  15. W. S. Yang, T. K. Kim, and R. N. Hill, "Performance Characteristics of Metal and Oxide Fuel Cores for a 1000 MWt Advanced Burner Reactor," Proc. of Workshop on Advanced Reactors With Innovative Fuels ARWIF-2008, Fukui, Japan, February 20-22, 2008.
  16. W. S. Yang, "Trends in Transmutation Performance and Safety Parameters Versus Conversion Ratio of Sodium- Cooled Recycle Reactors," Proc. of 10th Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation, Mito, Japan, October 6-10, 2008.
  17. J. E. Cahalan, M. A. Smith, R. N. Hill, and F. E. Dunn, "Physics and Safety Studies of a Low Conversion Ratio Sodium Cooled Fast Reactor," Proc. of PHYSOR 2004: The Physics of Fuel Cycles and Advanced Nuclear Systems: Global Developments, Chicago, Illinois, April 25-29, 2004.
  18. M. J. Grimstone, J. D. Tullett, and G. Rimpault, "Accurate Treatments of Fast Reactor Fuel Assembly Heterogeneity in the ECCO Cell Code," Proc. Intl. Conference on the Physics of Reactors, Marseilles, France, April 23-27, 1990.
  19. W. S. Yang, T. K. Kim, S. J. Kim, and C. H. Lee, Personal Communication, Argonne National Laboratory (2007).
  20. K. O. Ott and R. J. Neuhold, Introductory Nuclear Reactor Dynamics, American Nuclear Society, La Grange Park, Illinois (1985).
  21. H. H. Hummel and D. Okrent, Reactivity Coefficients in Large Fast Reactors, American Nuclear Society, La Grange Park, Illinois (1970).
  22. J. Graham, Fast Reactor Safety, Academic Press, New York, New York (1971).
  23. H. S. Khalil and R. N. Hill, "Evaluation of Liquid-Metal Reactor Design Options for Reduction of Sodium Void Worth," Nucl. Sci. Eng., 109 (1995).
  24. D. C. Wade and E. K. Fujita, "Trends versus Reactor Size of Passive Reactivity Shutdown and Control Performance," Nucl. Sci. Eng., 103, 182 (1989). https://doi.org/10.13182/NSE89-6
  25. W. S. Yang and T. A. Taiwo, "Status of Reactor Analysis Methods and Codes in the U.S.A," Proc. of PHYSOR 2004: The Physics of Fuel Cycles and Advanced Nuclear Systems: Global Developments, Chicago, Illinois, April 25-29, 2004.
  26. T. Takeda, "Neutronics Codes Currently Used in Japan for Fast and Thermal Reactor Applications," Proc. of PHYSOR 2004: The Physics of Fuel Cycles and Advanced Nuclear Systems: Global Developments, Chicago, Illinois, April 25-29, 2004.
  27. R. E. MacFarlane and D. W. Muir, "The NJOY Nuclear Data Processing System Version 91," LA-12740-M, Los Alamos National Laboratory (1994).
  28. L. B. Levitt, "The Probability Table Method for Treating Unresolved Resonances in Monte Carlo Criticality Calculations," Trans. Am. Nucl. Soc. 14, 648 (1971).
  29. R. D. Mosteller and R. C. Little, "Impact of MCNP Unresolved Resonance Probability-Table Treatment on Uranium and Plutonium Benchmarks," LA-UR-98-2943, Los Alamos National Laboratory (1998).
  30. "Cross Section Evaluation Working Group Benchmark Specifications," BNL-19302 (ENDF-202), Brookhaven National Laboratory (1974).
  31. "International Handbook of Evaluated Criticality Safety Benchmark Experiments," NEA/NSC/DOC(95)03, OECD Nuclear Energy Agency (1997).
  32. J. F. Briesmeister, ed., "MCNP - A General Monte Carlo N-Partile Transport Code, Version 4B," LA-12625-M, Los Alamos National Laboratory (1997).
  33. I. I. Bondarenko, et al, Group Constants for Nuclear Reactor Calculations, Consultants Bureau Enterprises, Inc., New York (1964).
  34. M. Segev, "A Theory of Resonance-Group Self-Shielding," Nucl. Sci. and Eng., 56, 72 (1975). https://doi.org/10.13182/NSE75-A26621
  35. C. R. Weisbin et al, "MINX: A Multigroup Interpretation of Nuclear X-Sections from ENDF/B," LA-6486-MS, Los Alamos Scientific Laboratory (1976).
  36. W. J. Davis, M. B. Yarbrough, and A. B. Bortz, "SPHINX, A One-Dimensional Diffusion and Transport Nuclear Cross Section Processing Code," WAPD-XS-3045-17, Westinghouse (1977).
  37. B. J. Toppel, H. Henryson II, and C. G. Stenberg, "ETOE-2/MC2-2/SDX Multi-group Cross-Section Processing," Conf-780334-5, Proc. of RSIC Seminar-Workshop on Multi-group Cross Sections, Oak Ridge, TN, March 1978.
  38. H. Henryson II, B. J. Toppel, and C. G. Stenberg, "MC2-2: A Code to Calculate Fast Neutron Spectra and Multigroup Cross Sections," ANL-8144, Argonne National Laboratory (1976).
  39. S. M. Bowman, D. F. Hollenbach, M. D. DeHart, B. T. Rearden, I. C. Gauld, S. Goluoglu, "An Overview of What's New in SCALE 5," Trans. Am. Nucl. Soc., 87 (2002).
  40. G. Rimpault, "Algorithmic Features of the ECCO Cell Code for Treating Heterogeneous Fast Reactor Subassemblies," Intl. Conf. on Mathematics and Computations, Reactor Physics, and Environmental Analyses, Portland, Oregon, April 30-May 4, 1990.
  41. M. N. Nikolaev, et al., "Method of Subgroups for Accounting of Resonance Structure of Cross-sections in Neutron Calculations," Atomn. Energ. 29, 11 (1970).
  42. D. E. Cullen, "Application of the Probability Table Method to Multi-group Calculations of Neutron Transport," Nucl. Sci. Eng. 55, 387 (1974). https://doi.org/10.13182/NSE74-3
  43. P. Ribon and J. M. Maillard, "Les Tables De Probabilite Applications Au Traitement Des Sections Efficaces Pour La Neutronique," Report CEA-N, NEACRP-L-294 (1986).
  44. L. C. Leal and R. N. Hwang, "Computation of Multipole Parameters Using Preliminary ENDF/B-VI Data", Trans. Am. Nucl. Soc., 62, 573 (1990).
  45. R. N. Hwang, "Generalized Pole Representation Revisited", Trans. Am. Nucl. Soc., 88, 491, (2003).
  46. R. N. Hwang, "A Rigorous Pole Representation of Multilevel Cross Sections and Its Practical Applications," Nucl. Sci. Eng., 96, 192 (1987). https://doi.org/10.13182/NSE87-A16381
  47. R. N. Hwang, "Efficient Methods for the Treatment of Resonance Integrals," Nucl. Sci. Eng., 52, 157 (1973). https://doi.org/10.13182/NSE73-A28186
  48. W. M. Stacey Jr., "Continuous Slowing Down Theory Applied to Fast-Reactor Assemblies," Nucl. Sci. Eng., 41, 381 (1970). https://doi.org/10.13182/NSE70-A19096
  49. F. L. Fillmore, "The CALHET-2 Heterogeneous Perturbation Theory Code and Application to ZPR-3-48," AI-69-13 (1969).
  50. R. E. Alcouffe, F. W. Brinkley, D. R. Marr, and R. D. O'Dell, "User's Guide for TWODANT: A Code Package for Two-Dimensional, Diffusion-Accelerated, Neutral-Particle Transport," LA-10049-M, Los Alamos National Laboratory (1990).
  51. W. S. Yang, M. A. Smith, C. H. Lee, A. Wollaber, D. Kaushik, and A. S. Mohamed, "Neutronics Modeling and Simulation of SHARP for Fast Reactor Analysis," Nuclear Engineering and Technology, 42, 520 (2010). https://doi.org/10.5516/NET.2010.42.5.520
  52. C. H. Lee and W. S. Yang, "MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis," ANL-NE-11-41, Argonne National Laboratory (2011).
  53. R. D. Lawrence, "The DIF3D Nodal Neutronics Option for Two- and Three-Dimensional Diffusion Theory Calculations in Hexagonal Geometry," ANL-83-1, Argonne National Laboratory (1983).
  54. T. J. Downar et al., "PARCS: Purdue Advanced Reactor Simulator," Proc. of PHYSOR 2002 ANS International Topical Meeting on the New Frontiers of Nuclear Technology, Seoul, Korea, October 7-10, 2002.
  55. T. Bahadir, S. Lindahl, and Palmtag, "SIMULATE-4 Multigroup Nodal Code with Microscopic Depletion Model," Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications, Avignon, France, September 12-15, 2005.
  56. T. A. Taiwo and H. S. Khalil, "DIF3D-K: A Nodal Kinetics Code for Solving the Time-Dependent Diffusion Equation," Proc. Int. Conf. Mathematics and Computations, Reactor Physics, and Environmental Analyses, Portland, Oregon, April 30-May 4, 1995.
  57. J. E. Cahalan, et al., "Advanced LMR Safety Analysis Capabilities in the SASSYS-1 and SAS4A Computer Codes," Proc. of the International Topical Meeting on Advanced Reactors Safety, Pittsburgh, PA (1994).
  58. T. A. Wareing, J. M. McGhee, and J. E. Morel, "ATTILA: A Three-Dimensional, Unstructured Tetrahedral Mesh Discrete Ordinates Transport Code," Trans. Am. Nucl. Soc., 75, 146 (1996).
  59. R. E. Alcouffe, R. S. Baker, J. A. Dahl, and S. A. Turner, "PARTISN Abstract," PHYSOR 2000 ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium, Pittsburgh, Pennsylvania, May 7-11, 2000.
  60. R. E. Alcouffe, R. S. Baker, F. W. Brinkley, D. R. Marr, R. D. O'Dell, and W. F. Walters, "DANTSYS: A Diffusion Accelerated Neutral Particle Transport Code System," LA-12969-M (1995).
  61. T. M. Evans, A. S. Stafford, R. N. Slaybaugh, and K. T. Clarno, "Denovo: A New Three-Dimensional Parallel Discrete Ordinates Code in SCALE," Nuclear Technology, 171, 171 (2010). https://doi.org/10.13182/NT171-171
  62. G. Palmiotti, E. E. Lewis, and C. B. Carrico, "VARIANT: VARIational Anisotropic Nodal Transport for Multidimensional Cartesian and Hexagonal Geometry Calculation," ANL-95/40, Argonne National Laboratory (1995).
  63. T. Taiwo, R. Ragland, G. Palmiotti, P. J. Finck, "Development of a Three-Dimensional Transport Kinetics Capability for LWR-MOX Analyses," Trans. Am. Nucl. Soc., 79, 298 (1999).
  64. M. Smith, D. Kaushik, A. Wollaber, W. S. Yang, and B. Smith, "New Neutronics Analysis Tool Development at Argonne National Laboratory," Proc. of International Conference on Fast Reactors and Related Fuel Cycles (FR09), Kyoto, Japan, December 7-11, 2009.
  65. M. A. Smith, A. Marin-Lafleche, W. S. Yang, D. Kaushik, and A. Siegel, "Method of Characteristics Development Targeting the High Performance Blue Gene/P Computer at Argonne National Laboratory," Proc. of International Conference on Mathematics, Computational Methods & Reactor Physics (M&C 2011), Rio de Janeiro, Brazil, May 8-12, 2011.
  66. A. Mohamed, W. S. Yang, M. A. Smith and C. H. Lee, "Analysis of Reaction Rate Distribution Measurements in ZPR-6 Assembly 7 Cores with $MC^{2}$-3/UNIC Code System," Proc. of International Conference on Mathematics, Computational Methods & Reactor Physics (M&C 2011), Rio de Janeiro, Brazil, May 8-12, 2011.
  67. R. M. Lell, J. A. Morman, R. W. Schaefer and R. D. McKnight, "ZPR-6 Assembly 7 High $^{240}Pu$ Core Experiments: A Fast Reactor Core with Mixed (Pu,U)-Oxide Fuel and a Central High $^{240}Pu$ Zone," International Handbook of Evaluated Reactor Physics Benchmark Experiments, NEA/NSC/DOC (2006)1, OECD-NEA (2009).
  68. R. M. Lell, "ZPR-6 Assembly 7 High $^{240}Pu$ Core: A Cylindrical Assembly with Mixed (Pu,U)-Oxide Fuel and a Central High $^{240}Pu$ Zone," International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)03, OECD-NEA (2009).
  69. F. B. Brown, et al. "MCNP5-1.51 Release Notes," LAUR-09-00384, Los Alamos National Laboratory (2009).
  70. S. Goluoglu, L. M. Petrie, Jr., M. E. Dunn, D. F. Hollenbach, and B. T. Rearden, "Monte Carlo Criticality Methods and Analysis Capabilities in SCALE," Nucl. Technol., 174, 214 (2011). https://doi.org/10.13182/NT10-124
  71. R. N. Blomquist, "VIM Continuous Energy Monte Carlo Transport Code," Proc. Intl. Conf. on Mathematics, Computations, Reactor Physics and Environmental Analysis, Portland, OR, April 30-May 4, 1995.
  72. L. S. Waters, ed., "MCNPX Users Manual (Version 2.1.5)," APT Program Report, Los Alamos National Laboratory (1999).
  73. R. E. Prael and H. Lichtenstein, "User Guide to LCS: The LAHET Code System," LA-UR-89-3014, Los Alamos National Laboratory (1989).
  74. B. J. Toppel, "A User's Guide to the REBUS-3 Fuel Cycle Analysis Capability," ANL-83-2, Argonne National Laboratory (1983).
  75. K. L. Derstine, "DIF3D: A Code to Solve One-, Two-, and Three-Dimensional Finite-Difference Diffusion Theory Problems," ANL-82-64, Argonne National Laboratory (1984).
  76. W. S. Yang and H. S. Khalil, "Analysis of the ATW Fuel Cycle Using the REBUS-3 Code System," Trans. Am. Nucl. Soc., 81, 277 (1999).
  77. W. S. Yang, J. C. Beitel, E. Hoffman, and J. A. Stillman, "Source Coupling Interface between MCNP-X and Deterministic Codes for ADS Analyses," Trans. Am. Nucl. Soc., 88, 592 (2003).
  78. W. H. Hannum, ed., "The Technology of the Integral Fast Reactor and its Associated Fuel Cycle," Prog. Nucl. Energy. (Special issue), 31 (1997).
  79. W. S. Yang, P. J. Finck, and H. Khalil, "Reconstruction of Pin Power and Burnup Characteristics from Nodal Calculations in Hexagonal Geometry," Nucl. Sci. Eng., 111, 21 (1992). https://doi.org/10.13182/NSE92-A23920
  80. "RSICC Computer Code Collection - ORIGEN 2.1, Isotope Generation and Depletion Code Matrix Exponential Method," CCC-371, Contributed by Oak Ridge National Laboratory (1999).
  81. W. B. Wilson et al., "Status of CINDER'90 Codes and Data," Proc. 4th Workshop on Simulating Accelerator Radiation Environments, Knoxville, Tennessee, September 13-16, 1998.
  82. I. C. Gauld, O. W. Hermann, and R. M. Westfall, "ORIGEN-S: SCALE System Module to Calculate Fuel Depletion, Actinide Transmutation, Fission Product Buildup and Decay, and Associated Radiation Source Terms," ORNL/NUREG/CSD-2/V2/R7, Oak Ridge National Laboratory (2002).
  83. I. C. Gauld, G. Radulescu, G. Ilas, B. D. Murphy, M. L. Williams, and D. Wiarda, "Isotopic Depletion and Decay Methods and Analysis Capabilities in SCALE," Nucl. Technol., 174, 169 (2011). https://doi.org/10.13182/NT11-3
  84. R. L. Moore, B. G. Schnitzler, C. A. Wemple, R. S. Babcock, And D. E. Wessol, "MOCUP: MCNP-ORIGEN2 Coupled Utility Program," INEL-95/0523, Idaho National Engineering Laboratory (1995).
  85. D. I. Poston, H. R. Trellue, "User's Manual Version 2.0, for MONTEBURNS Version 1.0," LA-UR-99-4999, Los Alamos National Laboratory (1999).
  86. Z. Xu and P. Hejzlar, "MCODE, Version 2.2: An MCNPORIGEN Depletion Program," MIT-NFC-TR-104, Massachusetts Institute of Technology (2008).
  87. A. Gandini, "Generalized Perturbation Theory for Nonlinear Systems from the Importance Conservation Principle," Nucl. Sci. Eng, 77, 316 (1981). https://doi.org/10.13182/NSE81-A19841
  88. M. L. Williams, "Development of Depletion Perturbation Theory for Coupled Neutron/Nuclide Fields," Nucl. Sci. Eng, 70, 20 (1979). https://doi.org/10.13182/NSE79-3
  89. G. Palmiotti, M. Salvatores, G. Aliberti, H. Hiruta, R. McKnight, P. Oblozinsky, W. S. Yang, "A Global Approach to the Physics Validation of Simulation Codes for Future Nuclear Systems," Ann. Nucl. Energy, 36, 355 (2009).
  90. C. H. Adams, Private Communication, Argonne National Laboratory (1975).
  91. W. S. Yang and T. J. Downar, "Generalized Perturbation Theory for Constant Power Core Depletion," Nucl. Sci. and Eng., 99, 353 (1988). https://doi.org/10.13182/NSE99-353
  92. W. S. Yang and T. J. Downar, "Depletion Perturbation Theory for the Constrained Equilibrium Cycle," Nucl. Sci. Eng., 102, 365-380 (1989). https://doi.org/10.13182/NSE89-A23648
  93. W. P. Poenitz and P. J. Collins, "Utilization of Experimental Integral Data for Adjustment and Uncertainty Evaluation of Reactor Design Quantities," NEACRP-L-307, Proc. NEACRP Specialists Meeting, Jackson Hole, WY (1988).
  94. P. J. Collins, S. E. Aumeier, and H. F. McFarlane, "Evaluation of Integral Measurements for the SP-100 Space Reactor," Proc. 1992 Topical Meeting on Advances in Reactor Physics, Charleston, SC (1992).
  95. W. S. Yang, G. Aliberti and R. D. McKnight, "Application of AFCI Covariance Data to Uncertainty Evaluation of Fast System Integral Parameters," J. Korean Physical Society, 59, 1288 (2011). https://doi.org/10.3938/jkps.59.1288
  96. P. Oblozinsky et al., "Progress on Nuclear Data Covariances: AFCI-1.2 Covariance Library," BNL-90897-2009, Brookhaven National Laboratory (2009).
  97. "Uncertainty and Target Accuracy Assessment for Innovative Systems Using Recent Covariance Data Evaluation," OECD/NEA Report No. 6410, OECD (2008).

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