Skip to main content

Core Materials

  • Chapter
  • First Online:
Fast Spectrum Reactors
  • 2786 Accesses

Abstract

The most hostile environment to be found in any nuclear reactor system is inside the core. Relative to a thermal reactor, the high flux, high burnup, and high temperature conditions encountered in a fast spectrum reactor place severe requirements on the materials selected for core design. Hence, considerable effort has been devoted to understanding and improving the performance of fuels and structural component candidates for fast spectrum reactor use.

This chapter is included to provide a more complete materials treatment of several of the general observations offered in Chapter 2 and of the design discussions included in Chapters 8, 9, and 10. However, because so much study has been given to the materials that comprise the primary building blocks of the fast spectrum reactor, it is not possible in an introductory text of this type to treat this subject with the degree of detail that a materials-oriented student would wish.

This is a preview of subscription content, log in via an institution to check access.

Access this chapter

Chapter
USD 29.95
Price excludes VAT (USA)
  • Available as PDF
  • Read on any device
  • Instant download
  • Own it forever
eBook
USD 219.00
Price excludes VAT (USA)
  • Available as EPUB and PDF
  • Read on any device
  • Instant download
  • Own it forever
Softcover Book
USD 279.99
Price excludes VAT (USA)
  • Compact, lightweight edition
  • Dispatched in 3 to 5 business days
  • Free shipping worldwide - see info
Hardcover Book
USD 379.99
Price excludes VAT (USA)
  • Durable hardcover edition
  • Dispatched in 3 to 5 business days
  • Free shipping worldwide - see info

Tax calculation will be finalised at checkout

Purchases are for personal use only

Institutional subscriptions

Notes

  1. 1.

    It is possible, of course, that such slowing down atoms could relocate in a vacancy position. This is called vacancy annihilation.

  2. 2.

    This normally causes densification.

  3. 3.

    A stoichiometric material contains an atomic ratio exactly equal to its chemical formula. Hypostoichiometric refers to a deficiency in the non-heavy metal constituent, and hyperstoichiometric refers to an excess in the non-heavy metal constituent.

  4. 4.

    Mixed oxide fuels undergo a volumetric increase of approximately 10% upon melting.

  5. 5.

    See Section 8.2 for a discussion of equiaxed and columnar grain development.

  6. 6.

    While most of the factors governing fission gas retention and redistribution are reasonably well understood, disagreement among ceramicists still remains over some of them [4]. Fuel temperature is only one of several governing factors.

  7. 7.

    Another difficulty of the carbide fuel fabrication process is the necessity to keep oxygen concentrations at very low levels. Uranium will preferably oxidize, and the presence of oxygen could produce unwanted UO2 as well as posing a fire hazard.

  8. 8.

    The term fuel “swelling” is somewhat ambiguous when applied to uranium metal because much of the observed growth over certain temperature ranges is due to the highly anisotropic behavior of uranium instead of fission product generation. However, the net effects of temperature and burnup are often referred to as swelling in the literature.

  9. 9.

    Fissium (Fs) is an equilibrium concentration of fission product elements left by the pyrometallurgical reprocessing cycle designed specifically for EBR-II. Nominal 5 wt.% fissium consists of 2.4 wt.% Mo, 1.9 wt.% Ru, 0.3 wt.% Rh, 0.2 wt.% Pd, 0.1 wt.% Zr, and 0.01 wt.% Nb.

  10. 10.

    Controlled venting could be purposely designed into the system.

  11. 11.

    Refer ahead to Fig. 11.19 for an explanation of the 0.2% offset concept.

  12. 12.

    This last requirement is a controversial point among materials specialists.

  13. 13.

    Inconel® is a registered trademark of International Nickel Company.

  14. 14.

    From page 28 of Ref. [29].

  15. 15.

    Considerable interest exists in venting the absorber pins in order to avoid the necessity of a plenum to contain the He buildup.

References

  1. D. R. Olander, “Fundamental Aspects of Nuclear Reactor Fuel Elements,” TID-26711-P1, Office of Public Affairs, U. S. ERDA, Washington, DC, 1976.

    Book  Google Scholar 

  2. S. Vana Varamban, “Estimation of solidus and liquidus temperature for UO2 and PuO2 psuedo binary system,” Proceedings of Sixteenth National Symposium on Thermal Analysis, THERMANS-2008, February 4–6, 2008, Kalpakkam, India. (Eds. Salil Varma, K.V. Govindan Kutty, S.K, Mukerjee, T. Gnanasekaran, Mrs. S.R. Bharadwaj, and V. Vengopal). Published by Scientific Information Resource Division, Bhabaha Atomic Research Centre, India, 2008, pp. 434–436.

    Google Scholar 

  3. Nuclear Energy Agency, OECD, “Accelerator-driven systems (ADS) and fast reactors (FR) in advanced nuclear fuel cycles.” NEA No. 3109, Paris, 2002.

    Google Scholar 

  4. E. H. Randklev, “Radial distribution of retained fission gas in irradiated mixed oxide fuel,” Trans. ANS, 28 (June 1978) 234–236.

    Google Scholar 

  5. M. Tourasse, M. Boidron, and B. Pasquet, “Fission products behaviour in Phénix fuel pins at high burnup,” J. Nucl. Mater., 188 (1992) 49.

    Article  Google Scholar 

  6. J. C. Melis, H. Plitz, and R. Thetford, “Highly irradiated fuel behaviour up to melting: JOG tests in CABRI,” J. Nucl. Mater., 204 (1993) 212.

    Article  Google Scholar 

  7. U. P. Nayak, A. Boltax, R. J. Skalka, and A. Biancheria, “An analytical comparison of the irradiation behavior of fast reactor carbide and oxide fuel pins,” Proceedings of the Topical Meeting Advanced LMFBR Fuels, Tucson, AZ, ERDA 4455, October 10–13, 1977, 537.

    Google Scholar 

  8. C. Ganguly and A. K. Sengupta, “Out-of-pile chemical compatibility of hyperstoichiometric (Pu0.7U0.3)C with stainless steel cladding and sodium coolant,” J. Nucl. Mater., 158 (1988) 159.

    Article  Google Scholar 

  9. B. Raj, “Plutonium and the Indian atomic energy programme,” J. Nucl. Mater., 385 (2009) 142.

    Article  Google Scholar 

  10. IAEA-TECDOC-1374, “Development status of metallic, dispersion and non-oxide advanced and alternative fuels for power and research reactors,” September, 2003.

    Google Scholar 

  11. T. N. Washburn and J. L. Scott, “Performance capability of advanced fuels for fast breeder reactors,” Proceedings of the Conference on Fast Reactor Fuel Element Technology, New Orleans, LA, April 13–15, 1971, 741–752.

    Google Scholar 

  12. J. A. L. Robertson, “Irradiation Effects in Nuclear Fuels,” Gordon and Breach, New York, NY, 1969, 223. (Copyright held by American Nuclear Society, LaGrange Park, IL.)

    Google Scholar 

  13. H. Blank, “Nonoxide Ceramic Nuclear Fuels,” from “Nuclear Materials,” Part I Volume 10 A, edited by B. R. T. Frost, appearing in “Materials Science and Technology, A Comprehensive Treatment,” edited by R. W. Cahn, P. Haasen, and E. J. Kramer, VCH Verlagsgesellschaft GmbH, Weinheim, 1994, 191.

    Google Scholar 

  14. H.J. Matzke, “Science of Advanced LMFBR Fuels,” chapter 4, “Physical properties” North-Holland, Amsterdam, 1986, 176.

    Google Scholar 

  15. W. F. Lyon, R. B. Baker, and R. D. Leggett, “Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor,” in LMR: A Decade of LMR Progress and Promise, Washington, DC, American Nuclear Society, La Grange Park, IL, November 11–15, 1990, 236–241.

    Google Scholar 

  16. A. A. Bauer, P. Cybulskis, and J. L. Green, “Mixed nitride fuel performance in EBR-II,” Proceedings of the Symposium on Advanced LMFBR Fuels, Tucson, AZ, October 10–13, 1977, pp. 299–312.

    Google Scholar 

  17. R. D. Leggett, R. K. Marshall, C. R. Hanu, and C. H. McGilton, “Achieving high exposure in metallic uranium fuel elements” Nucl. Appl. Tech., 9 (1970) 673.

    Google Scholar 

  18. C. W. Walter, G. H. Golden, and N. J. Olson, “U-Pu-Zr Metal Alloy: A Potential Fuel for LMFBR’s,” ANL 76-28, Argonne National Laboratory, Argonne, IL, November 1975.

    Google Scholar 

  19. G. L. Hofman, L. C. Walters, and T. H. Bauer, “ Metallic fast reactor fuels,” Progress in Nuclear Energy, 31 (1997) 83.

    Article  Google Scholar 

  20. R. G. Pahl, D. L. Porter, D. C. Crawford, and L. C. Walters, “Irradiation Behaviour of Metallic Fast Reactor Fuels,” ANL/CP-73323, Argonne National Laboratory, Argonne, IL, 1991.

    Google Scholar 

  21. B. R. Seidel and L. C. Walters, “Performance of metallic fuel in liquid-metal fast reactors,” ANS 1984, International meeting November 11–16, 1984, CONF-841105-2.

    Google Scholar 

  22. T. Ogata and T. Yokoo, “Development and validation of ALFUS: An irradiation behaviour analysis code for metallic fast reactor fuels,” Nucl. Tech., 128 (1999) 113.

    Google Scholar 

  23. R. G. Pahl, D. L. Porter, C. E. Lahm, and G. L. Hofman, “Experimental studies of U-Pu-Zr fast reactor fuel pins in EBR-II,” in AIME Meeting, Symposium on Irradiation Enhanced Materials Science, Chicago, IL, Fall 1988, CONF 8809202-2.

    Google Scholar 

  24. D. C. Crawford, D. L. Porter, and S. L. Hayes, “Fuels for sodium-cooled fast reactors: US perspective,” J. Nucl. Mater., 371 (2007) 202.

    Article  Google Scholar 

  25. A. Bauer, “Nitride fuels: properties and potential,” Reactor Tech., 15 (2) (1972).

    Google Scholar 

  26. V. S. Yemel'yanov and A. I. Yebstyukhin, “The Metallurgy of Nuclear Fuel-Properties and Principles of the Technology of Uranium, Thorium and Plutonium” (trans. Anne Foster), Pergamon Press, New York, NY, 1969.

    Google Scholar 

  27. J. H. Kittel, J. A. Horak, W. F. Murphy, and S. H. Paine, Effects of Irradiation on Thorium and Thorium-Uranium Alloys, ANL-5674, Argonne National Laboratory, Argonne, IL, 1963.

    Google Scholar 

  28. B. Blumenthal, J. E. Saneki, and D. E. Busch, “Thorium-Uranium-Plutonium Alloys as Potential Fast Power Reactor Fuels, Part II. Properties and Irradiation Behaviour of Thorium-Uranium-Plutonium Alloys,” ANL-7259, Argonne National Laboratory, Argonne, IL, 1969.

    Google Scholar 

  29. K. Wirtz, “Lectures on Fast Reactors,” Kernforschungszentrum, Karlsruhe, 1973 (Published by Gesellschaft fur Kernforschungszentrum).

    Google Scholar 

  30. J. K. Fink, M. G. Chasanov, and L. Leibowitz, “Thermophysical Properties of Thorium and Uranium System for Use in Reactor Safety Analysis,” ANL-CEN-RSD-77-1, Argonne National Laboratory, Argonne, IL, June 1977.

    Google Scholar 

  31. J. K. Fink, “Thermophysical properties of uranium dioxide,” J. Nucl. Mater., 279 (2000) 1.

    Article  Google Scholar 

  32. D. G. B. Martin, “The thermal expansion of solid UO2 and (U, Pu) mixed oxides — a review and recommendations,” J. Nucl. Mater., 152 (1988) 94.

    Article  Google Scholar 

  33. J. J. Carbajo, G. L. Yoder , S. G. Popov, and V. K. Ivanov, “A review of the thermophysical properties of MOX and UO2 fuels,” J. Nucl. Mater., 299 (2001) 181.

    Article  Google Scholar 

  34. Y. Philipponneau, “Thermal conductivity of (U, Pu)O2−x mixed oxide fuel,” J. Nucl. Mater., 188 (1992) 194.

    Article  Google Scholar 

  35. C. Duriez, J. P. Alessandri, T. Gervais, and Y. Philipponneau, “Thermal conductivity of hypostoichiometric low Pu content (U, Pu)O2−x mixed oxide,” J. Nucl. Mater., 277 (2000) 143.

    Article  Google Scholar 

  36. M. Inoue, “Thermal conductivity of uranium–plutonium oxide fuel for fast reactors,” J. Nucl. Mater., 282 (2000) 186.

    Article  Google Scholar 

  37. S. Majumdar, A. K. Sengupta, and H. S. Kamath, “Fabrication, characterization and property evaluation of mixed carbide fuels for a test fast breeder reactor,” J. Nucl. Mater., 352 (2006) 165.

    Article  Google Scholar 

  38. G. S. Was, “Fundamentals of Radiation Materials Science: Metals and Alloys,” Springer, Berlin, Heidelberg, 2007.

    Google Scholar 

  39. R. L. Fish and J. J. Holmes, “Tensile properties of annealed type 316 stainless steel after EBR-II irradiation,” J. Nucl. Mater., 46 (1973) 113.

    Article  Google Scholar 

  40. F. A. Garner, “Irradiation performance of cladding and structural steels in liquid metal reactors,” Material Science and Technology, edited by R. W. Cahn et al. Vol. 10A, VCH, Weinheim, 1994, 419–543.

    Google Scholar 

  41. C. W. Hunter, R. L. Fish, and J. J. Holmes, “Mechanical properties of unirradiated fast reactor cladding during simulated overpower transients,” Nucl. Tech., 27 (1975) 367.

    Google Scholar 

  42. E. E. Bloom and J. R. Weir, “Effect of neutron irradiation on the ductility of austenitic stainless steel,” Nucl. Tech., 16 (1972) 45.

    Google Scholar 

  43. R. L. Fish and C. W. Hunter, “Tensile properties of fast reactor irradiated type 304 stainless steel,” ASTM Symposium on the Effects of Radiation on Structural Materials, ASTM STP 611, 1976, 119.

    Google Scholar 

  44. T. Lauritzen, Stress-Rupture Behavior of Austenitic Steel Tubing: Influence of Cold Work and Effect of Surface Defects, USAEC Report GEAP-13897, 1972.

    Google Scholar 

  45. C. W. Hunter and G. D. Johnson, “Fuel adjacency effects on fast reactor cladding mechanical properties,” International Conference on Fast Breeder Reactor Fuel Performance, ANS/AIME, Monterey, CA, March 1979.

    Google Scholar 

  46. H. Farrar, C. W. Hunter, G. D. Johnson, and E. P. Lippincott, “Helium profiles across fast reactor fuel pin cladding” Trans. ANS, 23 (1976).

    Google Scholar 

  47. B. Raj, M. Vijayalakshmi, P. V. Sivaprasad, B. K. Panigrahi, and G. Amarendra, “Approaches to development of steel and manufacturing technology for fusion reactors,” Proceedings of the 29th Risø International Symposium on Energy Materials – “Advances in Characterization, Modelling and Application,” Editors Andersen, N. H.; Eldrup, M. Hansen, N. Juul Jensen, D. Nielsen, E. M. Nielsen, S. F. Sørensen, B. F. Pedersen, A. S. Vegge, T. West, S. S. Risø DTU, 2008.

    Google Scholar 

  48. S. Ukai, and T. Uwaba, “Swelling rate versus swelling correlation in 20% cold-worked 316 stainless steels,” J. Nucl. Mater., 317 (2003) 93.

    Article  Google Scholar 

  49. P. Dubuisson, A. Maillard, C. Delalande, D. Gilbon, and J. L. Seran, “The effect of phosphorus on the radiation-induced microstructure of stabilized austenitic stainless steel,” in Effects of Radiation on Materials, 15th International Symposium, STP 1125, ASTM, Philadelphia, PA, 1992, 995–1014.

    Google Scholar 

  50. F. A. Garner and D.S. Gelles, “Neutron-induced swelling of commercial alloys at very high exposures,” Effects of Radiation on Materials, 14th International Symposium (Vol II), STP 1046, ASTM, Philadelphia, 1990, 673–683.

    Google Scholar 

  51. J. L. Bates and M. K. Korenko, Updated Design Equations for Swelling of 20% CW AISI 316 Stainless Steel, HEDL-TME 78-3, Hanford Engineering Development Laboratory, Richland, WA, January 1978.

    Google Scholar 

  52. J. L. Straalsund, H. R. Brager, and J. J. Holmes, “Radiation-Induced Voids in Metals,” J. W. Corbett and L. C. Ianniello, eds. AEC Symposium Series No. 26, CONF-710601, 1972, 142.

    Google Scholar 

  53. J. J. Holmes, “Irradiation damage to cladding and structural materials – I,” Trans. ANS, 12 (1969) 117.

    Google Scholar 

  54. Y. Tateishi, “Development of long life FBR fuels with particular emphasis on cladding material improvement and fuel fabrication,” J. Nucl. Sci. Tech., 26 (1989) 132.

    Article  Google Scholar 

  55. N. Sekimura, T. Okita, and F. A. Garner, “Influence of carbon addition on neutron-induced void swelling of Fe–15Cr–16Ni–0.25Ti model alloy,” J. Nucl. Mater., 367–370 (2007) 897.

    Article  Google Scholar 

  56. C. Wassilew, K. Ehrlich, and J. J. Bergman, “Analysis of the in-reactor creep and rupture life behavior of stabilized austenitic stainless steels and nickel-based alloy hastellay-x,” Effects of Radiation on Materials, ASTM STP 956, ASTM, Philadelphia, 1987, 30.

    Google Scholar 

  57. C. H. Woo, “Irradiation creep due to elastodiffusion,” J. Nucl. Mater., 120 (1984) 55.

    Article  Google Scholar 

  58. K. Mansur, “Irradiation creep by climb-enabled glide of dislocations resulting from preferred absorption of point defects,” Phil. Mag. A, 39 (1979) 497.

    Article  Google Scholar 

  59. M. B. Toloczko and F. A. Garner, “Relationship between swelling and irradiation creep in cold-worked PCA stainless steel irradiated to ∼178 dpa at ∼400°C,” J. Nucl. Mater., 212–215 (1994) 509.

    Article  Google Scholar 

  60. B. Raj, “Materials science research for sodium cooled fast reactors,” Bull. Mater. Sci., 32 (2009) 271.

    Article  Google Scholar 

  61. L. Leibowitz, E. C. Chang, M. G. Chasanov, R. L. Gibby, C. Kim, A. C. Millunzi, and D. Stahl, “Properties for LMFBR Safety Analysis,” ANL-CEN-RS-76-1, Argonne National Laboratory, Argonne , IL, March 1976.

    Google Scholar 

  62. M. D. Mathew, “Towards developing an improved alloy D9 SS for clad and wrapper tubes of PFBR,” IGC News Letter, 75 (2008) 4.

    Google Scholar 

  63. P. Dubuisson, D. Gilbon, and J. L. Seran, “Microstructural evolution of ferritic-martensitic steels irradiated in the fast breeder reactor Phénix,” J. Nucl. Mater., 205 (1993) 178.

    Article  Google Scholar 

  64. R. L. Klueh and D. R. Harries, eds., “High Chromium Ferritic and Martensitic Steels for Nuclear Applications,” ASTM, Philadelphia, PA, 2001, 90.

    Google Scholar 

  65. A. Kohyama, A. Hishinuma, D. S. Gelles, R. L. Klueh, W. Dietz, and K. Elrich, “Low-activation ferritic and martensitic steels for fusion application,” J. Nucl. Mater., 233–237 (1996) 138.

    Article  Google Scholar 

  66. S. Ukai, S. Mizuta, M. Fujiwara, T. Okuda, and T. Kobayashi, “Development of 9Cr-ODS martensitic steel cladding for fuel pins by means of ferrite to austenite phase transformation,” J. Nucl. Sci. Tech., 39 (2002) 778.

    Article  Google Scholar 

  67. S. Ohtsuka, S. Ukai, and M. Fujiwara, “Nano-mesoscopic structural control in 9CrODS ferritic/martensitic steels,” J. Nucl. Mater., 351 (2006) 241.

    Article  Google Scholar 

  68. Handbook on Lead-bismuth Eutectic Alloy and Lead Properties, Materials Compatibility, Thermal-hydraulics and Technologies, 2007 Edition, OECD NEA No. 6195, ISBN 978-64-99002-9. Available on line at http://www.nea.fr/html/science/reports/2007/nea6195-handbook.htm.

  69. W. Hăfele, D. Faude, E. A. Fischer, and H. J. Laue, “Fast breeder reactors,” Annual Review of Nuclear Science, Annual Reviews, Inc., Palo Alto, CA, 1970.

    Google Scholar 

  70. J. K. Fink and L. Leibowitz, “Thermophysical Properties of Sodium,” ANL-CEN-RSD-79-1, Argonne National Laboratory, Argonne, IL, May 1979.

    Google Scholar 

  71. V. S. Bhise and C. F. Bonilla, “The experimental pressure and critical point of sodium,” Proceedings of the International Conference on Liquid Metal Technology in Energy Production, Seven Springs, PA, May 1977. (Also COO-3027-21, NTIS [1976].)

    Google Scholar 

  72. S. Das Gupta, “Experimental high temperature coefficients of compressibility and expansivity of liquid sodium and other related properties,” D.E.S. Dissertation with C. F. Bonilla, Dept. of Chemical Engineering and Applied Chemistry, Columbia University, Xerox-University Microfilms (1977). Also COO-3027-27, NTIS (1977).

    Google Scholar 

  73. G. H. Golden and J. D. Tokar, “Thermophysical Properties of Sodium,” ANL-7323, Argonne National Laboratory, Argonne, IL, 1967.

    Google Scholar 

  74. O. J. Foust, ed., “Sodium-NaK Engineering Handbook,” Vol. 1, Gordon and Breach, New York, NY, 1972, 23.

    Google Scholar 

  75. W. K. Anderson and J. S. Theilacker, eds., Neutron Absorber Materials for Reactor Control, U.S. Government Printing Office, Washington, DC, 1962.

    Google Scholar 

  76. Neutron Absorber Technology Staff, “A Compilation of Boron Carbide Design Support Data for LMFBR Control Elements,” HEDL-TME 75-19, Hanford Engineering Development Laboratory, Richland, WA, February 1975.

    Google Scholar 

  77. J. A. Basmajian and A. L. Pitner, “A correlation of boron carbide helium release in fast reactors,” Trans. ANS, 26 (1977) 174.

    Google Scholar 

  78. D. E. Mahagin and R. E. Dahl, “Nuclear Applications of Boron and the Borides,” HEDL-SA-713, Hanford Engineering Development Laboratory, Richland, WA, April 1974. (See also Ref. [70].)

    Google Scholar 

Download references

Author information

Authors and Affiliations

Authors

Corresponding author

Correspondence to Baldev Raj .

Editor information

Editors and Affiliations

Rights and permissions

Reprints and permissions

Copyright information

© 2012 Springer Science+Business Media, LLC

About this chapter

Cite this chapter

Raj, B. (2012). Core Materials. In: Waltar, A., Todd, D., Tsvetkov, P. (eds) Fast Spectrum Reactors. Springer, Boston, MA. https://doi.org/10.1007/978-1-4419-9572-8_11

Download citation

  • DOI: https://doi.org/10.1007/978-1-4419-9572-8_11

  • Published:

  • Publisher Name: Springer, Boston, MA

  • Print ISBN: 978-1-4419-9571-1

  • Online ISBN: 978-1-4419-9572-8

  • eBook Packages: EngineeringEngineering (R0)

Publish with us

Policies and ethics