Specific Zirconium Alloy Design Program Quarterly Progress Report: Fifth Quarter, April - June, 1963

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Description

A program is in progress for the design of a zirconium base alloy for steam service as nuclear fuel cladding. Thirty-one alloys selected for study of corrosion rate, hydriding rate and hydrogen embrittlement are in test. The corrosion testing of 1800 coupons to 3000 hours at at 300, 400, and 500 degrees C in refreshed steam has been completed. Statistical data analysis of the corrosion results are reported and alloys showing better corrosion performance at all test temperatures than that for Zircaloy-1 are discussed. Preliminary data for hydrogen uptake after long exposures at 400 and 500 degrees C are presented; … continued below

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x, 63 pages ; illustrations.

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

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A program is in progress for the design of a zirconium base alloy for steam service as nuclear fuel cladding. Thirty-one alloys selected for study of corrosion rate, hydriding rate and hydrogen embrittlement are in test. The corrosion testing of 1800 coupons to 3000 hours at at 300, 400, and 500 degrees C in refreshed steam has been completed. Statistical data analysis of the corrosion results are reported and alloys showing better corrosion performance at all test temperatures than that for Zircaloy-1 are discussed. Preliminary data for hydrogen uptake after long exposures at 400 and 500 degrees C are presented; the uptake for alloys showing the best corrosion performance is discussed. Post-corrosion mechanical property measurements are also reported along with the preliminary results of x-ray diffraction and metallographic studies relating to hydrogen embrittlement. A wide variation in resistance to embrittlement at a given hydrogen level was observed and can be tentatively correlated with original ductility, crystallographic texture, and hydride platelet orientation. The testing of a second round of ten alloys is also in progress. Studies concerning the mechanism of corrosion and hydriding in zirconium alloy are also reported. The results of recent neutron activation analyses of stripped corrosion films are presented. Oxygen diffusion through doped non-stoichiometric ZrO2 is now proceeding following earlier difficulties in sample preparation. Work on hydrogen overvoltage and electrochemical potential of inter-metallic phases was previously completed and reported.

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x, 63 pages ; illustrations.

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Digitized from microopaque cards.

Prepared under U.S. Atomic Energy Commission Contract AT(04-3)189 Project agreement 24 for the Joint U.S.-Euratom Research and Development Program.

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  • Report No.: GEAP-4284
  • Grant Number: AT(04-3)189 PA 24
  • OCLC: 897543796
  • SuDoc Number: Y 3.At 7:22/GEAP-4284
  • Accession or Local Control No: metadc1201359
  • Archival Resource Key: ark:/67531/metadc1201359

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Technical Report Archive and Image Library

The Technical Report Archive & Image Library (TRAIL) identifies, acquires, catalogs, digitizes and provides unrestricted access to U.S. government agency technical reports. The mission of TRAIL is to ensure preservation, discoverability, and persistent open access to government technical publications regardless of form or format.

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Imaged from microcard, these technical reports describe research performed for U.S. government agencies from the 1930s to the 1960s. The reports were provided by the Technical Report Archive and Image Library (TRAIL).

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  • July 1, 1963

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  • Aug. 15, 2019, 10:26 p.m.

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  • June 9, 2022, 7:33 p.m.

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Klepfer, H. H.; Jaech, John L.; Blood, R. E. & Douglass, D. L. (David Leslie), 1931-. Specific Zirconium Alloy Design Program Quarterly Progress Report: Fifth Quarter, April - June, 1963, report, July 1, 1963; San Jose, California.. (https://digital.library.unt.edu/ark:/67531/metadc1201359/: accessed June 9, 2023), University of North Texas Libraries, UNT Digital Library, https://digital.library.unt.edu; crediting UNT Libraries Government Documents Department.

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