37
Views
6
CrossRef citations to date
0
Altmetric
Technical Paper

Low-Temperature Rupture Behavior of Zircaloy-Clad Pressurized Water Reactor Spent Fuel Rods Under Dry Storage Conditions

&
Pages 107-123 | Published online: 10 May 2017

References

  • L. D. BLACKBURN, D. G. FARWICK, S. R. FIELD, L. A. JAMES and R. A. MOEN, “Maximum Allowable Temperature for Storage of Spent Nuclear Reactor Fuel,” HEDL-TME 78-37, Hanford Engineering Development Laboratory, Richland, Washington (May 1978).
  • F. R. LARSON and J. MILLER, “A Time-Temperature Relationship for Rupture and Creep Stresses,” Trans. ASME, 74, 765 (1952).
  • R. E. EINZIGER, S. D. ATKIN, D. E. STELLRECHT and V. PASUPATHI, “High Temperature Postirradiation Materials Performance of Spent Pressurized Water Reactor Fuel Rods Under Dry Storage Conditions,” Nucl. Technol., 57, 65 (1982).
  • R. S. KEMPER and D. L. ZIMMERMAN, “Neutron Irradiation Effects on the Tensile Properties of Zircaloy-2,” HW-52323, General Electric-Hanford Atomic Products Operation (Aug. 1957).
  • R. B. ADAMSON, “Irradiation Growth of Zircaloy,” Proc. 3rd Int. Conf. Zircaloy in the Nuclear Industry, Quebec City, Canada, August 10–12, 1976, ASTM STP 633, p. 326, A. L. LOWE and G. W. PARRY, Eds., American Society for Testing and Materials (1977).
  • K. PETTERSSON, “Rapid Stress Corrosion Crack Growth in Irradiated Zircaloy,” J. Nucl. Mater., 107, 117 (1982).
  • F. T. FUEILLO, S. PUSUHOTHAMAN and J. K. TIEN, “Understanding the Larson-Miller Parameter,” Scripta Meta., 11, 493 (1977).
  • E. HILLNER, “Corrosion and Hydriding Performance Evaluation of Three Zircaloy-2 Clad Fuel Assemblies After Continuous Exposure in PWR Cores 1 and 2 at Shipping-port, PA,” W-ARD-TM-1412, Westinghouse Advanced Reactors Division, Pittsburgh, Pennsylvania (Jan. 1980).
  • R. D. WATSON, “On the Oxidation of Zirconium Alloys in Air and the Dimensional Changes Associated with Oxidation,” AECL-3375, Atomic Energy of Canada, Chalk River, Ontario (June 1979).
  • D. G. BOASE and T. T. VANDERGRAAF, “The Canadian Spent Fuel Storage Canister: Some Materials Aspects,” Nucl. Technol., 32, 60 (Jan. 1977).
  • D. A. WOODFORD, “Creep Analysis of Zircaloy-4 and Its Application in the Prediction of Residual Stress-Relaxation,” J. Nucl. Mater., 79, 345 (1979).
  • H. E. ROSINGER and P. C. BERA, “Steady-State Creep of Zircaloy-4 Fuel Cladding from 940 to 1873 K,” J. Nucl. Mater., 82, 286 (1979).
  • D. L. HAGRMAN and G. A. REYMANN, “MATPRO-Version 11, A Handbook of Materials Properties for Use in the Analyses of Light Water Reactor Fuel Rod Behavior,” NUREG/CR-0497, U.S. Nuclear Regulatory Commission, Washington, D.C. (Feb. 1979).
  • K. R. MERCKX, “Calculational Procedures for Determining Creep Collapse of LWR Fuel Rods,” Nucl. Eng. Des., 31, 95 (1974).
  • P. J. PANKASKIE, “Irradiation Effects on the Mechanical Properties of Zirconium and Dilute Zirconium Alloys: A Review,” BN-SA-618, p. 47, Battelle Pacific Northwest Laboratories, Richland, Washington (July 1976).
  • D. E. STELLRECHT, V. PASUPATHI and J. C. KROGNESS, “Elevated Temperature Testing of Spent Nuclear Fuel Rods,” Trans. Am. Nucl. Soc., 34, 839 (June 1980)..
  • R. B. DAVIS, “Data Report for the Nondestructive Examination of Turkey Point Spent Fuel Assemblies B02, B03, B17, B41 and B43,” HEDL-TME 79-68, Hanford Engineering Development Laboratory, Richland, Washington (May 1980).
  • S. D. ATKIN, “Destructive Examination of 3-Cycle LWR Fuel Rods from Turkey Point Unit 3 for the Climax-Spent Fuel Test,” HEDL-TME 80-89, Hanford Engineering Development Laboratory, Richland, Washington (June 1980).
  • R. B. DAVIS, “Pre-Test Nondestructive Examination Data Summary Report on Turkey Point Spent Fuel Assemblies D01, D04, D06 for the Climax Spent Fuel Test,” HEDL-TME 80-83, Hanford Engineering Development Laboratory, Richland, Washington (June 1981).
  • R. B. DAVIS and V. PASUPATHI, “Data Summary Report for the Destructive Examination of Rods G7, G9, J8, 19 and H6 from Turkey Point Fuel Assembly B17,” HEDL-TME 80-85, Hanford Engineering Development Laboratory, Richland, Washington (Apr. 1981).
  • F. L. YAGGEE, R. F. MATTAS and L. A. NEIMARK, “Characterization of Irradiated Zircaloys: Susceptibility to Stress-Corrosion Cracking,” EPRI NP-1155, Electric Power Research Institute, Palo Alto, California (Sep. 1979).
  • D. CUBICIOTTI and R. L. JONES, “EPRI-NASA Cooperative Project on Stress-Corrosion Cracking in Zircaloys,” EPRI NP-717 Electric Power Research Institute, Palo Alto, California (Mar. 1978).
  • A. B. JOHNSON, E. R. GILBERT and R. J. GUENTHER, “Behavior of Spent Nuclear Fuel and Storage System Components in Dry Interim Storage,” PNL-4189, Pacific Northwest Laboratory, Richland, Washington (Aug. 1982).
  • A. TASOOJI, R. E. EINZIGER and A. K. MILLER, “Modelling of Zircaloy Stress Corrosion Cracking: Texture Effects and Dry Storage Spent Fuel Behavior,” Proc. 6th ASTM Symp. Zirconium in the Nuclear Industry, Vancouver, British Columbia, June 1982, ASTM STP-824, D. G. FRANKLIN and R. B. ADAMSON, Eds., American Society for Testing and Materials (Aug. 1984).
  • M. R. LOUTHAN, Jr. and R. P. MARSHALL, “Control of Hydride Orientation in Zircaloy,” J. Nucl. Mater., 9, 2, 170 (1963).
  • M. R. LOUTHAN, Jr. and C. L. ANGERMAN, “The Influence of Stress on the Hydride Habit Plane in Zircaloy-2,” Trans. TMS-AME, 236, 221 (1966).
  • “MATPRO Version 10, A Handbook of Materials Properties for Use in the Analyses of LWR Fuel Rod Behavior,” TREE NUREG-1180, EG&G Idaho, Inc., Idaho Falls, Idaho (Feb. 1978).
  • C. E. ELLIS, “Hydride Precipitates in Zirconium Alloys,” J. Nucl. Mater., 28, 129 (1968).
  • R. P. MARSHALL and M. R. LOUTHAN, Jr., “Tensile Properties of Zircaloy with Orientated Hydrides,” Trans. ASM, 56, 693 (1963).
  • E. D. HINDLE, “Effect of Circumferentially Aligned Hydrides on the Ductility of Zircaloy-2 Tubing Under Biaxial Stress,” J. Inst. Metals, 95, 359 (1967).
  • K. VIDEM, “Properties of Zirconium Base Cladding Materials –Corrosion and Hydrogen Pickup,” Nucl. Eng. Des., 21, 200 (1972).
  • R. P. MARSHALL, “Influence of Fabrication History on Stress-Oriented Hydrides in Zircaloy Tubing,” J. Nucl. Mater., 24, 34 (1967).
  • D. O. PICKMAN, “Properties of Zircaloy Cladding,” Nucl. Eng. Des., 21, 212 (1972).
  • D. HARDIE and M. W. SHANAHAN, “Stress Reorientation of Hydrides in Zr-2.5%Nb,” J. Nucl. Mater., 55 (1975).
  • G. P. WALTERS, “Hydride Morphology in Stressed Zirconium,” AERE R-5045, Atomic Energy Research Establishment, Harwell, U.K. (Sep. 1965).
  • G. W. PARRY, “Stress Reorientation of Hydride in Cold-Worked Zr-2.5%Nb Pressure Tubes,” AECL-2624, Atomic Energy of Canada, Chalk River, Ontario (Aug. 1966).
  • D. O. NORTHWOOD and U. KOSASIH, “Hydrides and Delayed Hydrogen Cracking in Zirconium and Its Alloy,” Int. Metall. Rev., 28, 2, 92 (1983).

Reprints and Corporate Permissions

Please note: Selecting permissions does not provide access to the full text of the article, please see our help page How do I view content?

To request a reprint or corporate permissions for this article, please click on the relevant link below:

Academic Permissions

Please note: Selecting permissions does not provide access to the full text of the article, please see our help page How do I view content?

Obtain permissions instantly via Rightslink by clicking on the button below:

If you are unable to obtain permissions via Rightslink, please complete and submit this Permissions form. For more information, please visit our Permissions help page.