In-reactor oxidation of zircaloy-4 under low water vapor pressures
Source abstract
Abstract
Abstract Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 °C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.
Source: openaire · Quality: candidate unverified · Source record ↗
Research context
OpenAlex topics: Nuclear Materials and Properties · Nuclear reactor physics and engineering · Radioactive element chemistry and processing
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References (17)
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1. R.A. Causey, D.F. Cowgill, R.H. Nilson, Review of the Oxidation Rate of Zirconium Alloys, 2005, SAND2005-6006.10.2172/876346 ↗
2. Cox · J. Nucl. Mater. · 2005 · 336 · 33110.1016/j.jnucmat.2004.09.029 ↗
3. Lefebvre · J. Nucl. Mater. · 1997 · 248 · 26810.1016/S0022-3115(97)00123-2 ↗
4. Bouineau · Nucl. Technol. · 2010 · 170 · 44410.13182/NT10-A10330 ↗
5. Zirconia powders and Zircaloy oxide films: tetragonal phase evolution during 400 C autoclave tests10.1016/0022-3115(95)00108-5 ↗
6. Dawson · J. Nucl. Mater. · 1968 · 25 · 17910.1016/0022-3115(68)90044-5 ↗
7. B. Cheng, H.A. Levin, R.B. Adamson, M.O. Marlow, V.L. Monroe, Development of a Sensitive and Reproducible Steam Test for Zircaloy Nodular Corrosion, Zirconium in the Nuclear Industry, 7th Int’l Symposium, ASTM STP939, R.B. Adamson, L.F.P. Van Swam (Eds.), ASTM Philadelphia, 1987, pp. 257–283.10.1520/STP28124S ↗
8. Iltis · J. Nucl. Mater. · 1995 · 224 · 10910.1016/0022-3115(95)00070-4 ↗
9. Iltis · J. Nucl. Mater. · 1995 · 224 · 12110.1016/0022-3115(95)00069-0 ↗
10. Tupin · J. Nucl. Mater. · 2003 · 317 · 13010.1016/S0022-3115(02)01704-X ↗
11. Westerman · J. Electrochem. Soc. · 1964 · 111 · 14010.1149/1.2426073 ↗
12. Standard Specification for Wrought Zirconium and Zirconium Alloy Seamless and Welded Tubes for Nuclear Service (Except Nuclear Fuel Cladding) ASTM B 353 ASTM International West Conshohocken, PA.
13. Johnson · 1987
14. E. Hillner, D.G. Franklin, J.D. Smee, The Corrosion of Zircaloy-Clad Fuel Assemblies in a Geologic Repository Environment, Bettis Atomic Power Lab., W. Mifflin, Report WAPD-T3173, 1994.
15. F. Garzarolli, W. Jung, H. Schoenfeld, A.M. Garde, G.W. Parry, P.G. Smerd, Waterside Corrosion of Zircaloy Fuel Rods, US Report, EPRI-NP-2789, Electric Power Research Inst., Palo Alto, CA, December, 1982.10.2172/6547121 ↗
16. Waterside corrosion of zirconium alloys in nuclear power plants, International Atomic Energy Agency, Vienna, Austria TECDOC-996, 1998.
17. K.J. Geelhood, C.E. Beyer, Hydrogen Pickup Models for Zircaloy-2, Zircaloy-4, M5TM, and ZIRLOTM, in: Proceedings of the 2011 Water Reactor Fuel Performance Meeting, September 11–14, Chengdu, China, 2011.