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Introduction and Overview

  • J. T. Adrian Roberts
Part of the Modern Perspectives in Energy book series (MOAC)

Abstract

Materials degradation in service represents one of the major technological factors that can limit the efficiency and viability of nuclear power. Extensive experience with commercial thermal reactors has demonstrated the need for improved understanding of materials phenomena (principally related to corrosion and irradiation) and better analytical procedures for transferring test information to the real problem. While fast breeder reactor experience is limited, it is already evident that currently used materials may not be sufficient for a completely economical breeder economy, and so advanced materials are being investigated for long-term application. The materials problems are much more diverse and severe for fusion reactors than for fission reactors, and no material considered to date possesses all the characteristics required for long-term operation in the fusion environment. Clearly, without intensive long-range development programs in this area, fusion may never transcend the engineering feasibility stage.

Keywords

Fatigue Crack Growth Austenitic Stainless Steel Fuel Assembly Steam Generator Linear Elastic Fracture Mechanic 
These keywords were added by machine and not by the authors. This process is experimental and the keywords may be updated as the learning algorithm improves.

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References

  1. 1.
    Fusion Power: Status and Options, Electric Power Research Institute, EPRI ER-510-SR (June 1977).Google Scholar
  2. 2.
    J. A. Maniscalco, D. H. Berwald and W. R. Meier, The Material Implications of Design and System Studies for Inertial Confinement Fusion Systems, paper presented at First Topical Meeting on Fusion Reactor Materials, American Nuclear Society, Bal Harbor, Florida (January 29–31, 1979).Google Scholar
  3. 3.
    Pool Type LMFBR Plant 1000 MW(e) Phase A Extension 2 Designs, Electric Power Research Institute Reports NP-1014-SY (GE), NP-1015-SY (AI), and NP-1016-SY (W) (1979).Google Scholar
  4. 4.
    M. E. Lapides, Nuclear Unit Productivity Analysis 1976 Update, Electric Power Research Institute, EPRI NP-559-SR (October 1977).Google Scholar
  5. 5.
    R. H. Koppe and E. A. Olson, Nuclear and Large Fossil Unit Operating Experience, Electric Power Research Institute, EPRI NP-1191 (September 1979).Google Scholar
  6. 6.
    J. James, Private communication, Electric Power Research Institute, (May 1979).Google Scholar
  7. 7.
    L. Minnick and M. Murphy, The Breeder: When and Why, Electr. Power Res. Inst. 7., No. 2, 6–11 (March 1976).Google Scholar
  8. 8.
    Section III, Division 1 Rules for Construction of Nuclear Power Plant Components, ASME Boiler and Pressure Vessel Code, American Society of Mechanical Engineers, New York (July 1977).Google Scholar
  9. 9.
    Code Case 1592, Class 1 Components in Elevated Temperature Service, Section III, ASME Boiler and Pressure Vessel Code, American Society for Mechanical Engineers, New York (November 1977).Google Scholar
  10. 10.
    Section XI-Rules for In-Service Inspection of Nuclear Power Plant Components, ASME Boiler and Pressure Vessel Code, American Society of Mechanical Engineers, New York (July 1977).Google Scholar
  11. 11.
    R. L. Jones, T. U. Marston, S. T. Oldberg, and K. E. Stahlkopf, Pressure Boundary Technology Program: Progress 1974 through 1978, Electric Power Research Institute, EPRI NP-1103-SR (March 1979).Google Scholar
  12. 12.
    Z. Zudans, M. M. Reddi, H. M. Fishman, and H. C. Tsai, Elastic-Plastic Creep Analysis of High-Temperature Nuclear Reactor Components, Nuc. Eng. Des. ,28, 414–445 (1974).CrossRefGoogle Scholar
  13. 13.
    Guidelines and Procedures for Design of Nuclear System Components at Elevated Temperatures, Reactor Development and Technology (RDT) Standard F9-ST (March 1974).Google Scholar
  14. 14.
    D. S. Griffin, Inelastic Structural Analysis: Design Implications and Experience, Nuc. Eng. Des. 51, 11–21 (1978).CrossRefGoogle Scholar
  15. 15.
    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. 97–113 (1976).Google Scholar
  16. 16.
    E. W. Hart, Constitutive Relations for the Nonelastic Deformation of Metals, J. Eng. Mater. Technol. ,193–202 (1976).Google Scholar
  17. 17.
    A. K. Miller, Development of the Materials Code, MATMOD (Constitutive Equations for Zircaloy), Electric Power Research Institute, EPRI NP-567, Project 456–1, Final Report (December 1977).Google Scholar
  18. 18.
    V. Kumar, S. Mukherjee, F. H. Huang, and Che-Yu Li, Deformation in Type 304 Stainless Steel, Electric Power Research Institute, EPRI NP-1276, Project 697–1, Final Report (December 1979).Google Scholar
  19. 19.
    E. W. Hart, C. Y. Li, H. Yamada, and G. L. Wire, Phenomenological Theory: A guide to Constitutive Relations and Fundamental Deformation Properties, in: Constitutive Equations in Plasticity ,A. S. Argon, ed., MIT Press, Cambridge, Massachusetts (1975), pp. 149–197.Google Scholar
  20. 20a.
    T. U. Marston, The EPRI Ductile Fracture Program, Proceedings of Seminar on Fracture Mechanics, April 2–6, 1979, ISPRA, Italy.Google Scholar
  21. 20b.
    P. C. Paris, CSNI Specialists Meeting on Plastic Tearing Instability, NUREG/CP-0010-CSNI Report No. 39, U.S. Nuclear Regulatory Commission, Washington, D.C. (October 1979).Google Scholar
  22. 20c.
    P. C. Paris and H. Tada, Further Results on the Subject of Tearing Instability. 1, NUREG/CR-1220, Vol. 1, U.S. Nuclear Regulatory Commission, Washington, D.C. (January 1980).Google Scholar
  23. 21.
    J. F. Knott, Fundamentals of Fracture Mechanics ,Butterworths, London (1973).Google Scholar
  24. 22.
    M. Reich and E. P. Esztergar, Compilation of References, Data Sources, and Analysis methods for LMFBR Primary Piping System Components, Brookhaven National Laboratory, BNL-NUREG 50650 (March 1977).Google Scholar
  25. 23.
    L. F. Coffin, S. S. Manson, A. E. Carden, L. K. Severud, and W. L. Greenstreet, Time-Dependent Fatigue of Structural Alloys, Oak Ridge National Laboratory, ORNL 5073 (January 1977).Google Scholar
  26. 24.
    H. Riedel, Creep Deformation at Crack Tips in Elastic-Viscoplastic Solids, Brown University Rep. No. MRL-E114 (June 1979).Google Scholar
  27. 25.
    EPRI Ductile Fracture Research Review Document (T. U. Marston, ed.), Electric Power Research Institute, EPRI NP-701-SR (February 1978).Google Scholar
  28. 25a.
    P. C. Paris, H. Tada, A. Zahoor and H. Ernst, Instability of the Tearing Mode of Elastic-Plastic Crack Growth, NUREG 0311, U.S. Nuclear Regulatory Commission, Washington, D.C. (August 1977).Google Scholar
  29. 26.
    P. Paris and F. Erogan, A Critical Analysis of Crack Propagation Laws, Trans. ASME ,528–534 (December 1963).Google Scholar
  30. 27.
    Boiler and Pressure Vessel Code, Section XI, Appendix A, Article A-4000, American Society of Mechanical Engineers, New York (1977).Google Scholar
  31. 28.
    L. A. James, Fatigue Crack Propagation Analysis of LMFBR Piping, In: Coolant Bound ary Integrity Considerations in Breeder Reactor Design ,Series PVP-PB-D27, R. H. Mallett and B. R. Nair, eds.), ASME (1978).Google Scholar
  32. 29.
    L. A. James, Fatigue Crack Propagation in Austenitic Stainless Steels, At. Energ. Rev. ,14, 37–85 (1976).Google Scholar
  33. 30.
    W. H. Bamford, Application of Corrosion Fatigue Crack Growth Rate Data to Integrity Analyses of Nuclear Reactor Vessels, paper presented at 3rd ASME National Congress on Pressure Vessels and Piping, San Francisco, California (June 1979).Google Scholar
  34. 31.
    S. R. Novak and S. T. Rolfe, Comparison of Fracture Mechanics and Nominal Stress Analysis in Stress Corrosion Cracking, Corrosion ,26, 121–130 (1970).Google Scholar
  35. 32.
    E. W. Hart, A Theory for Stable Crack Extension Rates in Ductile Materials, Report No. 4111, Materials Science Center, Cornell University (July 1979).Google Scholar
  36. 33.
    M. F. Ashby, First Report on Deformation-Mechanism Maps, Acta Metall ,20, 887–897 (1972).CrossRefGoogle Scholar
  37. 34.
    H. J. Frost and M. F. Ashby, Deformation Mechanism Maps for Pure Iron, Two Austenitic Stainless Steels, and a Low-Alloy Ferritic Steel, in: Fundamental Aspects of Structural Alloy Design ,R. I. Jaffee and B. A. Wilcox, eds.), Plenum Press, New York (1977), pp. 27–65.CrossRefGoogle Scholar
  38. 35.
    L. F. Coffin, Jr., The Multi-stage Nature of Fatigue: A Review, Met. Sci. ,68–72 (February 1977).Google Scholar
  39. 36.
    L. F. Coffin, Jr., A Note on Low-Cycle Fatigue Laws, J. Mater. ,6, 388–402 (1971).Google Scholar
  40. 37.
    S. S. Manson, Challenge to Unify Treatment of High-Temperature Fatigue-partisan proposal based on Strain-Range Partitioning, in: Symposium on Fatigue at Elevated Temperatures ,June 1972, Storrs, Connecticut, Am. Soc. Test. Mater. Spec. Tech. Publ. 520 (1972) pp. 744–782.Google Scholar
  41. 38.
    S. Majumdar and P. S. Maiya, A Unified and Mechanistic Approach to Creep-Fatigue Damage, in: Proceedings of the 2nd International Conference on Mechanical Behavior of Materials ,ICM-II, American Society of Metals, Metals Park, Ohio (1976), pp. 924–928.Google Scholar
  42. 39.
    A. E. Carden, Parametric Analysis of Fatigue Crack Growth, in: International Con ference on Creep and Fatigue in Elevated Temperature Applications ,Philadelphia (1973), Paper C324/73.Google Scholar
  43. 40.
    D. R. Olander, Fundamental Aspects of Nuclear Reactor Fuel Elements, Technical Information Center Energy Research and Development Administration, TID-26711-P1 (1976).Google Scholar
  44. 41.
    J. Gittus, Irradiation Effects in Crystalline Solids, Applied Science Publishers Ltd., London (1978).Google Scholar
  45. 42.
    F. A. Nichols, How Does One Predict and Measure Radiation Damage?, Nucl. Technol. ,40, 98–105 (1978).Google Scholar
  46. 43.
    G. L. Kulcinski, D. G. Doran, and M. A. Abdou, in: Properties of Reactor Structural Alloys after Neutron or Particle Irradiation ,Am. Soc. Test. Mater. Spec. Tech. Publ. 570 (1975), pp. 329–351.Google Scholar
  47. 44.
    S. K. Das and M. Kaminsky, Radiation Blistering of Structural Materials for Fusion Devices and Reactors, J. Nucl. Mater. ,53, 115–126 (1974).CrossRefGoogle Scholar
  48. 45.
    E. Hillner, Corrosion of Zirconium-Base Alloys-An Overview, Zirconium in the Nu clear Industry, Proceedings of the 3rd International Conference ,A. Lowe and G. Parry, eds., Am. Soc. Test. Mater. Spec. Tech. Publ. 633 (1977).Google Scholar
  49. 46.
    D. L. Douglass, The Metallurgy of Zirconium ,International Atomic Energy Agency, Vienna (1971).Google Scholar
  50. 47.
    R. Garnsey, Corrosion of PWR Steam Generators, Central Electricity Research Laboratories, England, RD/L/N4 79 (March 1979).Google Scholar
  51. 48.
    W. H. Layman, L. J. Martel, S. Green, G. Hetsroni, C. Shoemaker, and J. A. Mundis, Status of Steam Generators, paper presented to American Power Conference, April 1979, Chicago.Google Scholar

Copyright information

© Plenum Press, New York 1981

Authors and Affiliations

  • J. T. Adrian Roberts
    • 1
  1. 1.Electric Power Research InstitutePalo AltoUSA

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