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Part of the book series: Modern Perspectives in Energy ((MOAC))

Abstract

The extensive experience with the thermal reactor core materials (uranium dioxide, boron carbide, and stainless steel) has provided the basis for LMFBR core materials development. However, as noted in Chapter 1, the LMFBR fuel system must survive a much more hostile environment than the LWR fuel system with respect to temperature and fast neutron fluence. Typical fuel rod powers of 40 to 90 kW/m (compared to a maximum of ∼40 kW/m in LWRs) yield fast neutron fluences of ~2 × 1027 n/m2 (E > 0.1 MeV), or -90 dpa, with maximum cladding temperatures of the order of 920 K (Fig. 3-1). These more extreme conditions are manifested in more profound chemical and microstructural changes in the fuel and cladding; for instance, neither fission gas release approaching 100% nor void swelling of cladding are observed in LWR fuel rods, but they are normal occurrences in some LMFBR fuel systems and must therefore be accounted for in design. Similar allowances, although of a lesser nature, have had to be made in the LMFBR control rod design.

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References

  1. C. M. Cox, R. F. Hubert, and A. Biancheria, U.S. Experience in Irradiation Testing of Advanced Oxide Fuels, in: Proceedings of the American Nuclear Society Topical Meeting on Advanced LMFBR Fuels ,Tucson, Arizona (1977), pp. 136–148.

    Google Scholar 

  2. S. Kaplan, J. D. Stephen, R. M. Vijuk, and R. J. Jackson, Advanced Concepts for Fuel Assembly Design, in: Proceedings of the International Conference on Fast Breeder Reactor Fuel Performance ,Monterey, California, March 5–8, 1979, Pub. No. ISBN: 0-89448-105-3, pp. 758–776.

    Google Scholar 

  3. C. W. Weber, V. W. Lowery, A. Biancheria, and R. P. Omberg, The Performance of FBR Core Designs with Alternative Fuel, in: Proceedings of the International Confer ence on Fast Breeder Reactor Fuel Performance ,Monterey, California, March 5–8, 1979, Rep. No. ISBN: 0-89448-105-3, pp. 879–896.

    Google Scholar 

  4. Proceedings of the Conference on Fast Reactor Fuel Element Technology ,New Orleans, Louisiana, April 13–15, 1971, R. Farimaker, ed., American Nuclear Society.

    Google Scholar 

  5. Proceedings of the International Conference on Fast Breeder Reactor Fuel Perform ance ,Monterey, California, March 5–8, 1979, Pub. No. ISBN: 0-89448-105-3, Papers in Section IA, pp. 2–78.

    Google Scholar 

  6. H. Mikailoff, J. M. Dupouy, J. Ravier, M. Savineau, J. P. Pages, and J. M. Chaumont, Some Lessons Drawn from Operation of Phenix and Rhapsodie Cores, paper presented at European Nuclear Conference, Hamburg, Germany, March 6–11, 1979; abstract in Trans. Am. Nucl. Soc. ,31, 179–180 (1979).

    Google Scholar 

  7. Nuclear Systems and Materials Handbook ,Westinghouse Hanford Co., Richland, Washington, TID-26666, Vols. 1 and 2 [distribution limited to U.S. DOE approved recipients].

    Google Scholar 

  8. Proceedings of the International Conference on Fast Breeder Reactor Fuel Performance ,Monterey, California, March 5–8, 1979, Rep. No. ISBN: 0-89448-105-3, Section III, Fuel Design and Modeling, pp. 619–757.

    Google Scholar 

  9. D. Olander, Fundamental Aspects of Nuclear Reactor Fuel Elements, Technical Information Center, ERDA, TID-26711-P1 (1976).

    Book  Google Scholar 

  10. F. A. Nichols, Transport Phenomena in Nuclear Fuels under Severe Temperature Gradients, J. Nucl. Mater. ,84, 1–25 (1979).

    Article  Google Scholar 

  11. R. D. Leggett, E. N. Heck, P. J. Levine, and R. F. Hillbert, Steady-State Irradiation Behavior of Mixed Oxide Fuel Pins Irradiated in EBR-II, in: Proceedings of the Inter national Conference on Fast Breeder Reactor Fuel Performance ,Monterey, California, March 5–8, 1979, Rep. No. ISBN: 0-89448-105-3, pp. 2–15.

    Google Scholar 

  12. L. A. Lawrence, J. W. Weber, and J. L. De vary, Fuel-Cladding Chemical Interaction of Mixed Oxide Fuels, in: Proceedings of the International Conference on Fast Breeder Reactor Fuel Performance ,Monterey, California, March 5–8, 1979, Rep. No. ISBN: 0-89448-105-3, pp. 432–444.

    Google Scholar 

  13. D. C. Fee and C. E. Johnson, Fuel-Cladding Chemical Interaction in Uranium-Plutonium Oxide Fast Reactor Fuel Pins, J. Nucl. Mater. ,96(1981).

    Google Scholar 

  14. E. T. Weber, L. A. Lawrence, C. N. Wilson, and R. L. Gibby, In-Reactor Performance of Methods to Control Fuel-Cladding Chemical Interaction in: Proceedings of the In ternational Conference on Fast Breeder Reactor Fuel Performance ,Monterey, California, March 5–8, 1979, Rep. No. ISBN: 0-89448-105-3, pp. 445–459.

    Google Scholar 

  15. R. A. Karnesky, J. W. Jost, and I. Z. Stone, Cesium Migration in LMFBR Fuel Pins in: Proceedings of the International Conference on Fast Breeder Reactor Fuel Per formance ,Monterey, California, March 5–8, 1979, Rep. No. ISBN: 0-89448-105-3, pp. 343–352.

    Google Scholar 

  16. C. N. Wilson, R. L. Gibby, and E. T. Weber, Titanium Oxide Cesium Getters for Low O/M FTR Fuel Pins, paper presented at American Ceramic Society Basic Science and Nuclear Divisions Fall Meeting, October 14–17, 1979, New Orleans, Louisiana.

    Google Scholar 

  17. A. Boltax and A. Biancheria, Fuel-Cladding Mechanical Interaction Effects in Fast Reactor Mixed Oxide Fuel, paper presented at the IAEA International Working Group on Fast Reactors Technical Committee on Fuel and Cladding Interactions, Tokyo, Japan, February 21–25, 1977.

    Google Scholar 

  18. A. Biancheria, T. S. Roth, U. P. Nayak, and A. Boltax, Fuel-Cladding Mechanical Interaction in Fast Reactor Fuel Rods, in: Proceedings of the International Conference on Fast Breeder Reactor Fuel Performance ,Monterey, California, March 5–8, 1979, Rep. No. ISBN: 0-89448-105-3, pp. 513–535.

    Google Scholar 

  19. J. J. Holmes and J. L. Straalsund, Effects of Fast Reactor Exposure on the Mechanical Properties of Stainless Steels, in: Proceedings of the International Conference on Ra diation Effects in Breeder Reactor Structural Materials ,Scottsdale, Arizona, June 19–23, 1977, M. L. Bleiburg and J. W. Bennett, eds., American Institute of Mining Metallurgaical, and Petroleum Engineers, Inc. (1977), pp. 53–64.

    Google Scholar 

  20. J. L. Straalsund, R. L. Fish, and G. D. Johnson, Correlation of Transient Test Data with Conventional Mechanical Properties Data, Nucl. Technol. ,25, 531–540 (1975).

    Google Scholar 

  21. G. D. Johnson, J. L. Straalsund, and G. L. Wire, A New Approach to Stress-Rupture Data Correlation, Mater. Sci. Eng. ,28, 69–75 (1977).

    Google Scholar 

  22. A. K. Miller, An Inelastic Constitutive Model for Monotonic, Cyclic, and Creep Deformation. Part I. Equations Development and Analytical Procedures; Part II. Application to Type 304 Stainless Steel, J. Eng. Mater. Technol. ,98, 97–113 (1964).

    Article  Google Scholar 

  23. D. R. Harries, Neutron-Irradiation-Induced Embrittlement in Type 316 and Other Austenitic Steels and Alloys, J. Nucl. Mater. ,82. 2–21 (1979).

    Article  Google Scholar 

  24. M. L. Grossbeck, J. O. Stiegler, and J. J. Holmes, Effects of Irradiation on the Fracture Behavior of Austenitic Stainless Steels, in: Proceedings of the International Conference on Radiation Effects in Breeder Reactor Structural Materials ,June 19–23, 1977, M. L. Bleiburg and J. W. Bennett, eds., American Institute of Mining, Metallurgical, and Petroleum Engineers, Inc. (1977), pp. 53–64.

    Google Scholar 

  25. E. W. Hart and Che-Yu Li, Use of State Variables in the Description of Irradiation Creep and Deformation of Metals, in: Proceedings of the International Conference on Zirconium in the Nuclear Industry ,A. Lowe and G. Parry, eds., Am. Soc. Test. Mater. Spec. Tech. Publ. 633 (1977), pp. 315–325.

    Chapter  Google Scholar 

  26. G. Arnaud, A. Bernard, and P. Amman, Deformation of the Hexagonal Duct-Performance and Choice of Materials, Paper F14 in: Proceedings of the International Meeting on Irradiation Behavior of Metallic Materials for Fast Reactor Core Components, June 4–8, 1979, sponsored by CEA, Ajaccio, Corsica, France.

    Google Scholar 

  27. E. R. Gilbert, J. L. Straalsund, and G. L. Wire, Irradiation Creep Data in Support of LMFBR Core Design, J. Nucl. Mater ,65, 266–278 (1977).

    Article  Google Scholar 

  28. Proceedings of the International Conference on Radiation Effects in Breeder Reactor Structural Materials ,Scottsdale, Arizona, June 19–23, 1977, M. L. Bleiburg and J. W. Bennett, eds., American Institute of Mining, Metallurgical, and Petroleum Engineers (1977).

    Google Scholar 

  29. Proceedings of the International Conference on Radiation Effects in Breeder Reactor Structural Materials ,Scottsdale, Arizona, June 19–23, 1977, M. L. Bleiburg and J. W. Bennett, eds., American Institute of Mining, Metallurgical, and Petroleum Engineers (1977).

    Google Scholar 

  30. H. J. Bergmann, G. Knoblauch, D. Haas, and K. Herschbach, Examinations on Swelling and Irradiation Creep of the Austenitic Stainless Steel W-Nr. 1.4981CW, in: Proceedings of the International Meeting on Irradiation Behavior of Metallic Materials for Fast Reactor Core Components ,June 4–8, 1979, sponsored by CEA, Ajaccio, Corsica, France.

    Google Scholar 

  31. M. M. Paxton, B. A. Chin, E. R. Gilbert, and R. E. Nygren, Comparison of the In Reactor Creep of Selected Ferritic, Solid Solution Strengthened, and Precipitation Hardened Commercial Alloys, J. Nucl. Mater. ,80, 144–151 (1979).

    Article  Google Scholar 

  32. F. Rosa, P. S. Maiya, and R. W. Weeks, Effect of Grain Boundary Penetration of AISI Type 316 Stainless Steel by Cesium Oxides on Elevated Temperature Tensile Properties, in: Proceedings of the Symposium on Elevated Temperature Properties of Austenitic Stainless Steels ,A. O. Schaefer, ed., Metals Properties Council Inc., ASME, New York, (1974), pp. 97–112.

    Google Scholar 

  33. R. F. Hilbert, E. A. Aitken, and P. R. Pluta, High Burnup Performance of LMFBR Fuel Rods-Recent G. E. Experience, in: Progress in Nuclear Energy. Proceedings of the European Nuclear Conference, Vol. 3 (April 1975), pp. 526–529.

    Google Scholar 

  34. D. Lee, R. A. Rand, and G. G. Trantina, Crack Nucleation in 316 Stainless Steel Fuel Cladding, workshop paper presented at the 4th International Conference on Fracture, Waterloo, Canada, June 19–24, 1977, pp. 723–739.

    Google Scholar 

  35. C. Bagnall and D. G. Jacobs, Relationships for Corrosion of Type 316 Stainless Steel in Liquid Sodium, WARD-NA-3045-23 (1975), Westinghouse Electric Corp.

    Google Scholar 

  36. O. K. Chopra, J. Y. N. Wang, and K. Natesan, Review of Sodium Effects on Candidate Materials for Central Receiver Solar-Thermal Power Systems, Argonne National Laboratory Report, ANL-79-36 (July 1979).

    Book  Google Scholar 

  37. S. A. Shiels, A. R. Kerton, and R. P. Anatatmula, The In-Sodium Corrosion Behavior of Candidate Commercial Fuel Cladding and Duct Alloys, HEDL TME 77–71 (February 1978).

    Google Scholar 

  38. J. W. Weber, In-Reactor Corrosion Behavior of Stainless Steel Cladding in High-Temperature Sodium, presented at the International Conference on Liquid Metal Technology in Energy Production, Seven Springs, Pennsylvania, May 3–6, 1976.

    Google Scholar 

  39. J. W. Bennett, E. C. Norman, C. M. Cox, M. G. Adamson, A. Boltax, and J. H. Kittel, Advanced Alternate Breeder Fuels Testing in the U.S., in:Proceedings of the Inter national Conference on Fast Breeder Reactor Fuel Performance ,Monterey, California, March 5–8, 1979, Rep. No. ISBN: 0-89448-105-3, pp. 793–803.

    Google Scholar 

  40. E. Edmonds, W. Sloss, K. Q. Bagley, and W. Batey, Mixed Oxide Fuel Performance, in: Proceedings of the International Conference on Fast Breeder Reactor Fuel Per formance ,Monterey, California, March 5–8, 1979, Rep. No. ISBN: 0-89448-105-3, pp. 54–63.

    Google Scholar 

  41. A. L. Lotts, compiler, Fast Breeder Reactor Oxide Fuels Development-Final Report, Oak Ridge National Laboratory Publication No. 4901 (November 1973).

    Google Scholar 

  42. L. A. Neimark, Argonne National Laboratory Progress Reports ANL-RDP35, December 1974, p. 5.12; ANL-RDP38 ,March 1975, p. 5.17; ANL-RDP41 ,June 1975, p. 5.6.

    Google Scholar 

  43. W. G. Johnston, J. H. Rosolowski, A. M. Turkalo, and T. Lauritzen, An Experimental Survey of Swelling in Commercial Fe-Cr-Ni Alloys Bombarded with 5 MeV Ni Ion, J. Nucl. Mater. ,54, 24–40 (1974).

    Article  Google Scholar 

  44. D. R. Harries, The UKAEA Fast Reactor Project Research and Development Program on Fuel Element Cladding and Sub-Assembly Wrapper Materials, in: Proceedings of the International Conference on Radiation Effects in Breeder Reactor Structural Ma-42.terials ,Scottsdale, Arizona, June 19–23, 1977, M. L. Bleiberg and J. W. Bennett, eds., Met. Soc. AIME (1977), pp. 27–40.

    Google Scholar 

  45. J. J. Laidler, J. J. Holmes, and J. W. Bennet, U.S. Programs on Reference and Advanced Cladding/Duct Materials, in: Proceedings of the International Conference on Radiation Effects in Breeder Reactor Structural Materials ,June 19–23, 1977, Scottsdale, Arizona, M. L. Bleiburg and J. W. Bennett, eds., by Met. Soc. AIME (1977), pp. 41–52.

    Google Scholar 

  46. J. Lehman, J. M. Dupouy, R. Broudeur, J. L. Boutard, and A. Maillard, Irradiation Creep of 316 and 316 Ti Stainless Steels, in: Proceedings of the International Meeting on Irradiation Behavior of Metallic Materials for Fast Reactor Core Components ,June 4–8, 1979, sponsored by CEA, Ajaccio, Corsica, France.

    Google Scholar 

  47. E. A. Little and D. A. Stow, Void Swelling in Irons and Ferritic Steels. II. An Experimental Survey of Materials Irradiated in a Fast Reactor, J. Nucl. Mater. ,87, 25–39 (1979).

    Article  Google Scholar 

  48. J. Erler, A. Maillard, G. Brun, J. Lehmann, and J. M. Dupouy, The Behavior of Ferritic Steels under Irradiation with Fast Neutrons, in: Proceedings of the International Meeting on Irradiation Behavior of Metallic Materials for Fast Reactor Core Components ,June 4–8, 1979, sponsored by CEA, Ajaccio, Corsica, France.

    Google Scholar 

  49. J. J. Huet, A. Delbrassine, P. Van Asbroeck, and W. Vandermeulen, Radiation Effects in Ferritic Steels in: Proceedings of the International Conference on Radiation Effects in Breeder Reactor Structural Materials ,June 19–23, 1977, Scottsdale, Arizona, M. L. Bleiburg and J. W. Bennett, eds., Met. Soc. AIME (1977), pp. 357–366.

    Google Scholar 

  50. E. A. Little, Void Swelling in Irons and Ferritic Steels. I. Mechanisms of Swelling Suppression, J. Nucl. Mater ,87, 11–24 (1979).

    Article  Google Scholar 

  51. G. W. Cunningham, Materials Development for Advanced Reactors, Nucl. Technol. ,28, 301–304 (1976).

    Google Scholar 

  52. J. M. Dupouy, French Program on LMFBR Cladding Materials Development, in: Proceedings of the International Conference on Radiation Effects in Breeder Reactor Struc tural Materials ,June 19–23, 1977, Scottsdale, Arizona, M. L. Bleiburg and J. W. Bennett, eds., Met. Soc. AIME (1977), pp. 1–12.

    Google Scholar 

  53. J. M. Kendall, LMFBR Steam Cycles-Is Efficiency the Ultimate Goal?, paper presented at 1979 Annual Meeting of ANS, Atlanta, Georgia, June 3–7, 1979; summary in: Trans. Am. Nucl. Soc. ,32, 564–565 (1979).

    Google Scholar 

  54. J. O. Barner, T. W. Latimer, J. F. Kerrisk, R. L. Petty, and J. L. Green, Advanced Carbide Fuels-U.S. Experience, in: Proceedings of the Topical Meeting on Advanced LMFBR Fuels ,Tucson, 1977, American Nuclear Society and Energy Research and Development Administration, pp. 268–298.

    Google Scholar 

  55. T. W. Latimer, R. L. Petty, J. F. Kerrisk, N. S. DeMuth, P. J. Levine, and A. Boltax, Irradiation Performance of Helium-Bonded Uranium-Plutonium Carbide Fuel Elements, in: Proceedings of the International Conference on Fast Breeder Reactor Fuel Performance ,Monterey, California, March 5–8, 1979, Pub. No. ISBN: 0-89448-105-3, pp. 816–826.

    Google Scholar 

  56. J. F. Kerrisk, N. S. DeMuth, R. L. Petty, T. W. Latimer, J. A. Vitti, and L. J. Jones, Design and Performance of Sodium-Bonded Uranium-Plutonium Carbide Fuels, in: Proceedings of the International Conference on Fast Breeder Reactor Fuel Performance ,Monterey, California, March 5–8, 1979, Rep. No. ISBN: 0-89448-105-3, pp. 804–815.

    Google Scholar 

  57. R. W. Stratton and L. Smith, The Irradiation Behavior of Spherepac Carbide Fuel, in: Proceedings of the Topical Meeting on Advanced LMFBR Fuels ,Tucson, Arizona, 1977, American Nuclear Society and Energy Research and Development Administration, pp. 79–85.

    Google Scholar 

  58. K. Q. Bagley, W. Batey, R. Paris, W. M. Sloss, and G. P. Snape, U.K. Irradiation

    Google Scholar 

  59. Experience Relevant to Advanced Carbide Fuel Concepts for LMFBRs, in: Proceedings of the Topical Meeting on Advanced LMFBR Fuels ,Tucson, Arizona, 1977, American Nuclear Society and Energy Research and Development Administration, pp. 313–325.

    Google Scholar 

  60. A. A. Bauer, P. Cybulskis, and J. L. Green, Mixed-Nitride Fuel Performance in EBR II, in: Proceedings of the Topical Meeting on Advanced LMFBR Fuels ,Tucson, Arizona, 1977, American Nuclear Society and Energy Research and Development Administration, pp. 299–312.

    Google Scholar 

  61. A. A. Bauer, P. Cybulskis, N. S. DeMuth, and R. L. Petty, He-and Na-Bonded Mixed Nitrite Fuel Performance, in: Proceedings of the International Conference on Fast Breeder Reactor Fuel Performance ,Monterey, California, American Nuclear Society, Rep. No. ISBN: 0-89448-105-3, pp. 827–841.

    Google Scholar 

  62. R. A. Murgatroyd and B. T. Kelly, Technology and Assessment of Neutron Absorbing Materials, At. Energ. Rev. ,15, 3–74 (1977).

    Google Scholar 

  63. Compilation of B4C Design Data for LMFBRs, HEDL-TME 75–19 (1975).

    Google Scholar 

  64. D. E. Mahagin and R. E. Dahl, Nuclear Applications of Boron and the Borides, in: Boron and Refractory Borides ,V. I. Matkovich, ed., Springer Verlag, Heidelburg(1977), Chapter VII, pp. 613–632.

    Chapter  Google Scholar 

  65. J. A. Basimajin, A. L. Pitner, D. E. Mahagin, H. C. F. Ripfel, and D. E. Baker, Irradiation Effects in Boron Carbide Pellets Irradiated in Fast Neutron Spectra, Nucl. Technol. ,16, 238–248 (1972).

    Google Scholar 

  66. G. W. Hollenburg, J. L. Jackson, and J. A. Basmajian, In-Reactor Measurement of Neutron Absorber Performance, Nucl. Technol. ,49, 92–101 (June 1980).

    Google Scholar 

  67. IAEA Specialist Meeting on Absorbing Materials and Control Rods for Fast Reactors, Dimitrovgrad, U.S.S.R., June 1973, Summary Report edited by R. E. Dahl and J. W. Bennett, HEDL TME 73–91 (1973).

    Google Scholar 

  68. G. W. Hollenburg, B. Mastel, and J. A. Basmajian, Helium Bubbles in Irradiated Boron Carbide, J. Am. Ceram. Soc. ,63 (7-8), 376–380 (1980).

    Article  Google Scholar 

  69. C. E. Beyer and G. W. Hollenberg, Physically Based Model for Helium Release from Irradiated Boron Carbode, paper presented at American Ceramic Society Basic Science and Nuclear Divisions Fall Meeting, October 14–17, 1979, New Orleans, Louisiana.

    Google Scholar 

  70. G. W. Hollenburg and W. V. Cummings, Effect of Fast Neutron Irradiation on the Structure of Boron Carbide, J. Am. Ceram. Soc. ,60, 520–525 (1977).

    Article  Google Scholar 

  71. T. Inoue, T. Onchi, and H. Koyama, Irradiation Effects of Boron Carbide used as Control Rod Elements in Fast Breeder Reactors, J. Nucl. Mater. ,74, 114–122 (1978).

    Article  Google Scholar 

  72. A. E. Pasto and M. M. Martin, Eu2O3: Properties and Irradiation Behavior, ORNL 5291, Oak Ridge National Laboratory (1977).

    Book  Google Scholar 

  73. A. E. Pasto and V. J. Tennery, Results of BICM-2 Irradiation Test of Eu2O3 in EBR II, ORNL-5345, Oak Ridge National Laboratory (December 1977).

    Google Scholar 

  74. M. Levenson and E. L. Zebroski, A Fast Breeder System Concept: A Diversion-Resistant Fuel Cycle, Nucl. Eng. Des. ,51, 119–132 (1979).

    Article  Google Scholar 

  75. DOE Alternate Breeder Fuel Development Program, Reports of Task Groups (May 1977).

    Google Scholar 

  76. W. N. Bishop, Postirradiation Examination of Thoria-Uranium Fuel Rods, Babcock &Wilcox Publication No. 3809-7 (October 1969).

    Google Scholar 

  77. J. T. A. Roberts, Ceramic Utilization in the Nuclear Industry: Current Status and Future Trends, Part II, Powder Metall. Int. ,11, 72–80 (1979).

    Google Scholar 

  78. J. H. Kittel, D. L. Johnson, W. N. Beck, and J. A. Horak, Metallic Fuel Systems for Alternative Breeder Fuel Cycles, in: Proceedings of the International Conference on Fast Breeder Reactor Fuel Performance ,Monterey, California, 1979, Am. Nucl. Soc. Pub. ISBN: 0-89448-105-3, pp. 925–934.

    Google Scholar 

  79. J. H. Kittel and L. C. Walters, Development and Performance of Metal Fuel Elements for Fast Breeder Reactors, paper presented at the European Nuclear Conference, May 6–11, 1979, Hamburg, West Germany; abstract in Trans. Am. Nucl. Soc. ,31, 177–178 (1979).

    Google Scholar 

  80. P. S. K. Lam, R. B. Turski, and W. P. Barthold, Performance of U-Pu-Zr Metal Fuel in 1000 MWe LMFBRs, in: Proceedings of the International Conference on Fast Breeder Reactor Fuel Performance ,Monterey, California, 1979, Am. Nucl. Soc. Pub. No. ISBN: 0-89448-105-3, pp. 935–953.

    Google Scholar 

  81. C. M. Walter, G. H. Golden, and N. J. Olsen, U-Pu-Zr Metal Alloy: A Potential Fuel for LMFBRs, Argonne National Laboratory Report, ANL-76-28 (November 1975).

    Google Scholar 

  82. G. J. Fischer and R. J. Cerbone, The Fast Mixed Spectrum Reactor: Interim Report, Initial Feasibility Study, BNL-50976, Brookhaven National Laboratory (1979).

    Google Scholar 

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Roberts, J.T.A. (1981). LMFBR Core Materials. In: Structural Materials in Nuclear Power Systems. Modern Perspectives in Energy. Springer, Boston, MA. https://doi.org/10.1007/978-1-4684-7194-6_3

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