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Oxidation of Zircaloy-4 during in situ proton irradiation and corrosion in PWR primary water

Published online by Cambridge University Press:  16 March 2015

Peng Wang  and
Gary S. Was
Peng Wang*
Affiliation:
Nuclear Engineering and Radiological Sciences Department, University of Michigan, Ann Arbor, Michigan 48109, USA
Gary S. Was
Affiliation:
Nuclear Engineering and Radiological Sciences Department, University of Michigan, Ann Arbor, Michigan 48109, USA
*
a)Address all correspondence to this author. e-mail: [email protected]
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Abstract

The kinetics and morphology of oxides formed during in situ proton irradiation–corrosion experiment were analyzed. Experiments were conducted in 320 °C water with 3 wt ppm H2, while irradiated by a 3.2-MeV proton beam at a current density of 2 µA/cm2 producing a damage rate at 4.4 × 10−7 dpa/s. The resulting oxide was compared with reference samples corroded in an autoclave, and literature data found on in-reactor formed oxide. The corrosion rate of the sample irradiated in situ was 10 times faster than the in-pile corrosion rate. The cracked and porous irradiated oxide consisted of monoclinic equiaxed grains of zirconia with a preferential orientation of the oxide grains. Second phase particles (SPPs) consumed by the oxidation front were rapidly oxidized, but no SPPs were amorphized or dissolved in the metal matrix of the irradiated sample.

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Articles
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Journal of Materials Research , Volume 30 , Issue 9: Focus Issue: Characterization and Modeling of Radiation Damage on Materials: State of the Art, Challenges, and Protocols , 14 May 2015 , pp. 1335 - 1348
Copyright
Copyright © Materials Research Society 2015 

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Footnotes

Contributing Editor: Joel Ribis

References

REFERENCES

Garzarolli, F., Broy, Y., and Busch, R.A.: Comparison of the long-time corrosion behavior of certain Zr alloys in PWR, BWR, and laboratory tests. Zirconium in the Nuclear Industry: Eleventh International Symposium, ASTM STP 1295, Bradley, E.R. and Sabol, G.P. eds. (American Society for Testing and Materials, West Conshohocken, PA, 1996); p. 850.CrossRefGoogle Scholar
Iltis, X., Lefebvre, F., and Lemaignan, C.: Microstructural study of oxide layers formed on Zircaloy-4 in autoclave and in reactor, Part II: Impact of the chemical evolution of intermetallic precipitates on their zirconia environment. J. Nucl. Mater. 224, 121 (1995).CrossRefGoogle Scholar
Ghosal, S., Palit, G.C., and De, P.K.: Corrosion of zirconium alloys in nuclear applications—a review. Miner. Process. Extr. Metall. Rev. 22, 519 (2001).Google Scholar
Asher, R.C., Davies, D., and Kirstein, T.B.A.: The corrosion of some zirconium alloys under radiation in moist carbon dioxide-air mixtures. J. Nucl. Mater. 49(2), 189 (1973).CrossRefGoogle Scholar
Bradhurst, D.H., Shirvington, P.J., and Heuer, P.M.: The effects of radiation and oxygen on the aqueous oxidation of zirconium and its alloys at 290°C. J. Nucl. Mater. 46(1), 53 (1973).CrossRefGoogle Scholar
Johnson, A.B. Jr. and Irvin, J.E.: Radiation-Enhanced Oxidation of Zircaloy-2 in Ph-10 NH4OH: Technical Report BNWL-463 Battelle Pacific Northwest Laboratories, Richland, Washington, 1967.Google Scholar
Cox, B.: Some thoughts on the mechanisms of in-reactor corrosion of zirconium alloys. J. Nucl. Mater. 336(2–3), 331 (2005).Google Scholar
Woo, O.T., McDougall, G.M., Hutcheon, R.M., Urbanic, V.F., Griffiths, M., and Coleman, C.E.: Corrosion of electron-irradiated Zr-2.5Nb and zircalpy-2. Zirconium in the Nuclear Industry: Twelfth International Symposium, ASTM STP 1354, Sabol, G.P. and Moan, G.D., eds.; American Society for Testing and Materials: West Conshohocken, PA, 2000; p. 709.Google Scholar
Cheng, B.C., Kruger, R.M., and Adamson, R.B.: Corrosion behavior of irradiated Zircaloy. Zirconium in the Nuclear Industry: Tenth International Symposium, ASTM STP 1245, Garde, A.M. and Bradley, E.R., eds.; American Society for Testing and Materials: Philadelphia, PA, 1994; p. 400.Google Scholar
LaVerne, J.A.: H2 formation from the radiolysis of liquid water with zirconia. J. Phys. Chem. B 109, 5395 (2005).CrossRefGoogle ScholarPubMed
LaVerne, J.A. and Tandon, L.: H2 production in the radiolysis of water on CeO2 and ZrO2 . J. Phys. Chem. B 106(2), 380 (2002).CrossRefGoogle Scholar
LaVerne, J.A. and Tandon, L.: H2 production in the radiolysis of water on UO2 and other oxides. J. Phys. Chem. B 107(49), 13623 (2003).Google Scholar
Cox, B. and Fidleris, V.: Enhanced low-temperature oxidation of zirconium alloys under irradiation. Zirconium in the Nuclear Industry: Eighth International Symposium, ASTM STP 1023, Van Swam, L.F.P. and Eucken, C.M., eds.; American Society for Testing and Materials: Philadelphia, PA, 1989; p. 245.Google Scholar
Franklin, D. and Lang, P.M.: Zirconium-alloy corrosion: A review based on an international atomic energy agency (IAEA) meeting. Zirconium in the Nuclear Industry: Ninth International Symposium, ASTM STP 1132, Eucken, C.M. and Garde, A.M., eds.; American Society for Testing and Materials: Philadelphia, PA, 1991; p. 3.Google Scholar
Pêcheur, D., Lefebvre, F., Motta, A.T., Lemaignan, C., and Charquet, D.: Oxidation of intermetallic precipitates in Zircaloy-4: Impact of irradiation. Zirconium in the Nuclear Industry: Tenth International Symposium, ASTM STP 1245, Garde, A.M. and Bradley, E.R., eds.; American Society for Testing and Materials: Philadelphia, PA, 1994; p. 687.CrossRefGoogle Scholar
Fidleris, V., Tucker, R.P., and Adamson, R.B.: An overview of microstructural and experimental factors that affect the irradiation growth behavior of zirconium alloys. Zirconium in the Nuclear Industry: Seventh International Symposium, ASTM STP 939, Adamson, R.B. and Van Swam, L.F.P., eds.; American Society for Testing and Materials: Philadelphia, PA, 1987; p. 49.Google Scholar
Bradley, E.R. and Perkins, R.A.: Characterization of Zircaloy corrosion films by analytical transmission electron microscopy. In Proceedings of IAEA Technical Committee Meeting on Fundamental Aspect of Corrosion of Zirconium-Base Alloys in Water Reactor Environments, IAEA, Vienna, IWGFPT/34, Vol. 9, (1990).Google Scholar
Raiman, S., Flick, A., Toader, O., Wang, P., Samad, N.A., Jiao, Z., and Was, G.S.: A facility for studying irradiation accelerated corrosion in high temperature water. J. Nucl. Mater. 451(1–3), 40 (2004).Google Scholar
Ziegler, J.F.: The Stopping and Range of Ions in Matter (SRIM). 2013. Available from http://www.srim.org.Google Scholar
ASTM E521. Standard Practice for Neutron Radiation Damage Simulation by Charged-Particle Irradiation. (American Society for Testing and Materials, Philadelphia, 1996).Google Scholar
JCPDS Database, International Center for Diffraction Data, 2002.Google Scholar
Barberis, P., Merle-Méjean, T., and Quintard, P.: On Raman spectroscopy of zirconium oxide films. J. Nucl. Mater. 246(2–3), 232 (1997).CrossRefGoogle Scholar
Motta, A., Erwin, K.T., Delaire, O., Birtcher, R.C., Chu, T., Maser, J., Mancini, D.C., and Lai, B.: Synchrotron radiation study of second phase partials and alloying elements in zirconium alloys. Zirconium in the Nuclear Industry: Thirteenth International Symposium, ASTM STOP 1423, Moan, G.D. and Rudling, P., eds.; ASTM International: West Conshohocken, PA, 2002; p. 59.Google Scholar
Bouineau, V., Bénier, G., Pêcheur, D., Thomazet, J., Ambard, A., and Blat, M.: Analysis of waterside corrosion kinetics of Zircaloy-4 fuel cladding in French PWRs. Nucl. Technol. 170(3), 444 (2010).Google Scholar
Bryner, J.S.: The cyclic nature of corrosion of Zircaloy-4 in 633 K water. J. Nucl. Mater. 82(1), 84 (1979).Google Scholar
Hillner, E., Franklin, D.G., and Smee, J.D.: Long-term corrosion of Zircaloy before and after irradiation. J. Nucl. Mater. 278(2–3), 334 (2000).Google Scholar
Kido, T., Kanasugi, K., Sugano, M., and Komatsu, K.: PWR zircaloy cladding corrosion behavior: Quantitative analyses. J. Nucl. Mater. 248, 281 (1997).Google Scholar
Allison, C.M., Berna, G.A., Chambers, R., Coryell, E.W., Davis, K.L., Hagrman, D.L., Hagrman, D.T., Hampton, N.L., Hohorst, J.K., Mason, R.E., McComas, M.L., McNeil, K.A., Miller, R.L., Olsen, C.S., Reymann, G.A., and Siefken, L.J.. “SCDAP/RELAP5/MOD3.1 Code Manual Volume IV: MATPRO – A Library of Materials Properties for Light-Water-Reactor Accident Analysis”. D.T. Hagrman, NUREG/CR-6150, EGG-2720, Volume IV. 1993, p. 4–234.Google Scholar
Allen, T.R., Cole, J.I., and Kenik, A.: Radiation-induced segregation and void swelling in 304 stainless steel. Effects of Radiation on Materials: Twentieth International Symposium, ASTM STP 1405, Rosinski, S.T., Grossbeck, M.L., Allen, T.R. and Kumar, A.S., eds.; American Society for Testing and Materials: West Conshohocken, PA, 2001; p. 428.Google Scholar
Van Duysen, J.C., Todeschini, P., and Zacharie, G.: Effect of neutron irradiations at temperature below 500°C on the properties of cold worked 316 stainless steels: A review. Effects of Radiation Materials: Sixteenth International Symposium ASTM STP 1175, Kumar, A.S., Gelles, D.S., Nanstad, R. and Little, E.A., eds.; American Society for testing and materials: Philadelphia, PA, 1993; p. 747.Google Scholar
Iltis, X., Lefebvre, F., and Lemaignan, C.: Microstructure evolutions and iron redistribution in zircaloy oxide layers: Comparative effects of neutron irradiation flux and irradiation damages. Zirconium in the Nuclear Industry: Eleventh International Symposium, ASTM STP 1295. (American Society for Testing and Materials, West Conshohocken, PA, 1996); p. 242.Google Scholar
Spitznagel, J.A., Fleischer, L.R., and Choyke, W.J.: The effects of ion bombardment on the thin film oxidation behavior of Zircaloy-4 and Zr-1.0Nb. Proceeding of International Conference on Application of Ion Beams to Metals, Albuquerque, NM, Vol. 87, (1973).Google Scholar
Corrosion of zirconium alloys in nuclear power plants, IAEA, Vienna, IAEA-TECDOC-684, ISSN 1011-4289 (1993).Google Scholar
Lefebvre, F. and Lemaignan, C.: Irradiation effects on corrosion of zirconium alloy claddings. J. Nucl. Mater. 248, 268 (1997).Google Scholar
Gong, W., Zhang, H., Qiao, Y., Xian, H., Ni, X., Li, Z., and Wang, X.: Grain morphology and crystal structure of pre-transition oxides formed on Zircaloy-4. Corros. Sci. 74, 323 (2013).Google Scholar
Ni, N., Lozano-Perez, S., Sykes, J.M., Smith, G.D.W., and Grovenor, C.R.M.: Focused ion beam sectioning for the 3D characterization of cracking in oxide scales formed on commercial ZIRLOTM alloys during corrosion in high temperature pressurized water. Corros. Sci. 53(12), 4073 (2011).Google Scholar
Parise, M., Foerch, R., and Cailletaud, G.: Coupling between diffusion and mechanics during the oxidation of Zircaloy. J. Phys. IV France 9, Pr9–311 (1999).Google Scholar
Pilling, N.B. and Bedworth, R.E.: The oxidation of metals at high temperatures. J. Inst. Met 29, 529 (1923).Google Scholar
Parise, M., Sicardy, O., and Cailletaud, G.: Modeling of the mechanical behavior of the metal-oxide system during Zr alloy oxidation. J. Nucl. Mater. 256(1), 35 (1998).Google Scholar
Bossis, P., Lelièvre, G., Barberis, P., Iltis, X., and Lefebvre, F.: Multi-scale characterization of the metal-oxide interface of zirconium alloys. Zirconium in the Nuclear Industry: Twelfth International Symposium, ASTM STP 1354, Sabol, G.P. and Moan, G.D., eds.; American Society for Testing and Materials: West Conshohocken, PA, 2000; p. 918.Google Scholar
Proff, C., Abolhassani, S., and Lemaignan, C.: Oxidation behavior of zirconium alloys and their precipitates–A mechanistic study. J. Nucl. Mater. 432(1–3), 222 (2013).Google Scholar
Garzarolli, F., Seidel, H., Tricot, R., and Gros, J.P.: Oxide growth mechanism on zirconium alloys. Zirconium in the Nuclear Industry: Ninth International Symposium, ASTM STP 1132, Eucken, C.M. and Garde, A.M., eds.; American Society for Testing and Materials: Philadelphia, PA, 1991; p. 395.Google Scholar
Pêcheur, D., Lefebvre, F., Motta, A.T., Lemaignan, C., and Wadier, J.F.: Precipitate evolution in the Zircaloy-4 oxide layer. J. Nucl. Mater. 189(3), 318 (1992).Google Scholar
Hood, G.M. and Schultz, R.J.: Diffusion of 3D transition elements in alpha-Zr and zirconium alloys. Zirconium in the Nuclear Industry: Eighth International Symposium, ASTM STP 1023, Van Swam, L.F.P. and Eucken, C.M., eds.; American Society for Testing and Materials: Philadelphia, PA, 1989; p. 435.CrossRefGoogle Scholar
Baek, J-H. and Jeong, Y-H.: Depletion of Fe and Cr within precipitates during Zircaloy-4 oxidation. J. Nucl. Mater. 304(2–3), 107 (2002).Google Scholar
Murty, K.L.: Materials’ ageing and degradation in light water rectors, Mechanisms and management (Woodhead Publishing Limited, Cambridge, UK, 2013); p. 164.Google Scholar