Specific Zirconium Alloy Design Program Quarterly Progress Report: Sixth Quarter, July - September, 1963

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Summary: Fundamental studies in support of the alloy design work are complete except for the experimental determination of the diffusion of oxygen in alloy-doped non-stoichiometric ZrO2. Over 100 oxidation runs have now been made on samples of ZrO2 doped with 1 mole percent of the oxides of Al, Y, Fe, Cr, and Ni. The first round testing of 31 alloys is now essentially complete. Analysis of the steam corrosion rate and hydriding raw data taken at 300, 400, and 500 degrees C indicates that Zr-Cr and Zr-Cu-Fe alloys show the most promise for development for service in steam over the … continued below

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ii, 36 pages ; illustrations.

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Klepfer, H. H.; Jaech, John L.; Douglass, D. L. (David Leslie), 1931-; Blood, R. E. & Perrine, H. E. October 1, 1963.

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  • Vallecitos Atomic Laboratory
    Publisher Info: General Electric Company. Vallecitos Atomic Laboratory. Atomic Power Equipment Department.
    Place of Publication: San Jose, California

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Summary: Fundamental studies in support of the alloy design work are complete except for the experimental determination of the diffusion of oxygen in alloy-doped non-stoichiometric ZrO2. Over 100 oxidation runs have now been made on samples of ZrO2 doped with 1 mole percent of the oxides of Al, Y, Fe, Cr, and Ni. The first round testing of 31 alloys is now essentially complete. Analysis of the steam corrosion rate and hydriding raw data taken at 300, 400, and 500 degrees C indicates that Zr-Cr and Zr-Cu-Fe alloys show the most promise for development for service in steam over the entire temperature range 300-500 degrees C. Maximum resistance to corrosion hydrogen embrittlement requires high initial ductility and thus low, perhaps less than ~2.5 a/o total alloy content. For any composition, susceptibility to hydrogen embrittlement depends on crystallographic texture of the component; under certain circumstances hydrogen embrittlement may be high anisotropic. The second-round testing of 10 selected Zr-Cr and Zr-Cu base alloys is now about 50% complete. Three alternate fabrication schedules were evaluated; and the preliminary results indicate that the Zr-Cu alloy tested is less sensitive to heat treatment than is the Zr-Cr alloy tested. Raising the final alpha annealing temperature from 565 degrees C to 788 degrees C gives better over-all corrosion and hydrating performance for both the Zr-Cr and Zr-Cu alloy tested. Beryllium additions to Zr-Cr or Zr-Cu do not appear to be advantageous. Nickel additions to Zr-Cu do not give an over-all improvement. Nickel additions to Zr-Cu give about the same improvement over Zr-Cu as did iron additions to the Zr-Cu in the first-round test.

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ii, 36 pages ; illustrations.

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U.S. Atomic Energy Commission Contract AT(04-3)-189 Project Agreement 24.

Includes bibliographic references (page 35).

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  • Report No.: GEAP-4368
  • Grant Number: AT(04-3)-189 PA 24
  • SuDoc Number: Y 3.At 7:22/GEAP-4368
  • Accession or Local Control No: metadc1201386
  • Archival Resource Key: ark:/67531/metadc1201386

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  • October 1, 1963

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  • Aug. 15, 2019, 10:26 p.m.

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  • June 25, 2024, 1:27 a.m.

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Klepfer, H. H.; Jaech, John L.; Douglass, D. L. (David Leslie), 1931-; Blood, R. E. & Perrine, H. E. Specific Zirconium Alloy Design Program Quarterly Progress Report: Sixth Quarter, July - September, 1963, report, October 1, 1963; San Jose, California. (https://digital.library.unt.edu/ark:/67531/metadc1201386/: accessed January 12, 2026), University of North Texas Libraries, UNT Digital Library, https://digital.library.unt.edu; crediting UNT Libraries Government Documents Department.

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