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Technical Paper

UO2 Pellet-Stack Shortening in a Boiling Water Reactor

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Pages 103-108 | Published online: 13 May 2017

References

  • F. LIST and P. KNUDSEN, “Pellet Stack Shortening and Cladding Elongation of Irradiated UO2-Zr Fuel Pins,” Risø Report No. 256, 45-52, Danish Atomic Energy Commission, Risø, Roskilde, Denmark (1972).
  • G. KJAERHEIM and E. ROLSTAD, “In-Core Study of the Mechanical Interaction between Fuel and Cladding,” Nucl. Appl. Technol., 7, 347 (1969).
  • J. B. AINSCOUGH, “Some Limiting Aspects of UO2 Performance,” TRG Report 1937(S), United Kingdom Atomic Energy Authority (1969).
  • J. B. AINSCOUGH, United Kingdom Atomic Energy Authority, Personal Communication (1971).
  • F. A. NICHOLS, “On the Mechanisms of Irradiation Creep in Zirconium-Base Alloys,” J. Nucl. Mater., 37, 59 (1970).
  • D. S. WOOD and B. WATKINS, “A Creep Limit Approach to the Design of Zircaloy-2 Reactor Pressure Tubes at 275°C,” J. Nucl. Mater., 41, 327 (1971).
  • J. E. HARBOTTLE, “The Temperature and Neutron Dose Dependence of Irradiation Growth in Zircaloy 2,” Irradiation Effects on Structural Alloys for Nuclear Reactor Applications, ASTM STP 484, pp. 287–299, American Society for Testing Materials (1970).
  • B. LUSTMAN, “Fuel Clad Design Basis for Thermal Reactors,” WAPD-T-1939, Bettis Atomic Power Laboratory (1966).
  • J. VEEDER, “Thermo-Elastic Expansion of Finite Cylinders,” AECL-2260, Atomic Energy of Canada Ltd. (1967).

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