journal article·2015-01

Assessment of oxygen diffusion coefficients by studying high-temperature oxidation behaviour of Zr1Nb fuel cladding in the temperature range of 1100–1300 °C

M. Négyesi, T. Chmela, T. Veselský, J. Krejčí, L. Novotný, A. Přibyl, O. Bláhová, J. Burda, J. Siegl, V. Vrtílková

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Abstract

Abstract The paper deals with high-temperature steam oxidation behaviour of Zr1Nb fuel cladding. First of all, comprehensive experimental program was conducted to provide sufficient experimental data, such as the thicknesses of evolved phase layers and the overall weight gain kinetics, as well as the oxygen concentration and nanohardness values at phase boundaries. Afterwards, oxygen diffusion coefficients in the oxide, in the α-Zr(O) layer, in the double-phase (α + β)-Zr region, and in the β-phase region have been estimated based on the experimental data employing analytical solution of the multiphase moving boundary problem, assuming the equilibrium conditions being fulfilled at the interface boundaries. Eventually, the determined oxygen diffusion coefficients served as input into the in-house numerical code, which was designed to predict the high-temperature oxidation behaviour of Zr1Nb fuel cladding. Very good agreement has been achieved between the numerical calculations and the experimental data.
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Research context

OpenAlex topics: Nuclear Materials and Properties · High Temperature Alloys and Creep · Intermetallics and Advanced Alloy Properties

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References (28)

Show source-provided references
1. Chung · Nucl. Eng. Technol. · 2005 · 37 · 327
2. C. Grandjean, G. Hache, A state-of-the-art review of past programmes devoted to fuel behaviour under loss-of-coolant conditions. Part 3. Cladding Oxidation. Resistance to Quench and Post-Quench Loads, DPAM/SEMCA 2008–093, IRSN, June 2008.
3. H.M. Chung, T.F. Kassner, Embrittlement criteria for Zircaloy fuel cladding applicable to accident situations in light-water reactors, NUREG/CR-1344, ANL-79-48, Argonne National Laboratory, January 1980.10.2172/5496633 ↗
4. D.O. Hobson, Ductile-brittle behavior of Zircaloy fuel cladding, Proc. ANS Topical Mtg. on Water Reactor Safety, Salt Lake City, March 26, 1973, pp. 274–288.10.2172/4569664 ↗
5. Pawel · J. Nucl. Mater. · 1974 · 50 · 24710.1016/0022-3115(74)90095-6 ↗
6. A. Sawatzky, Proposed criterion for the oxygen embrittlement of Zircaloy-4 fuel cladding, in: Proc. 4th Symp. On Zirconium in the Nuclear Industry, Stratford-on-Avon, UK, June 27–29, 1978.10.1520/STP36697S ↗
7. J.-C. Brachet et al., Hydrogen content, pre oxidation and cooling scenario influences on post-quench mechanical properties of Zy-4 and M5(TM) alloys in LOCA conditions – relationship with the post-quench microstructure, in: 15th International Symposium on Zirconium in the Nuclear Industry, June 24–28, 2007, Sun River, Oregon, USA, J. ASTM Int. vol. 5, No. 5, Paper ID JAI101116, 2008.10.1520/STP48132S ↗
8. Negyesi · Nucl. Eng. Des. · 2013 · 261 · 26010.1016/j.nucengdes.2012.09.033 ↗
9. G. Hache, H.M. Chung, The history of LOCA embrittlement criteria. Topical Meeting on LOCA Fuel Safety Criteria, Aix-en-Provence, March 22–23, 2001, pp. 37–64.
10. V. Vrtilkova, L. Novotny, R. Doucha, J. Vesely, An approach to the alternative LOCA embrittlement criterion, in: Proceedings of Meeting of OECD Nuclear Energy Agency Committee on the Safety of Nuclear Installations Special Expert Group on Fuel Safety Margins (SEGFSM), May 25–26, 2004, Argonne, Illinois, USA.
11. Portier · J. ASTM Int. · 2005 · 2 · 10310.1520/JAI12468 ↗
12. Corvalán-Moya · J. Nucl. Mater. · 2010 · 400 · 19610.1016/j.jnucmat.2010.03.004 ↗
13. Chung · J. Nucl. Mater. · 1979 · 84 · 32710.1016/0022-3115(79)90172-7 ↗
14. Pawel · J. Electrochem. Soc. · 1979 · 126 · 111110.1149/1.2129228 ↗
15. Ma · J. Nucl. Mater. · 2008 · 377 · 35910.1016/j.jnucmat.2008.03.012 ↗
16. Négyesi · J. Nucl. Mater. · 2011 · 416 · 29810.1016/j.jnucmat.2011.06.015 ↗
17. Nikulin · J. Nucl. Mater. · 2011 · 418 · 110.1016/j.jnucmat.2011.07.017 ↗
18. J.C. Brachet et al., Quantification of the α(O) and prior-β phase fractions and their oxygen contents in high temperature (HT) oxidised Zr alloys (Zy-4, M5™), in: Proceedings of SEGFSM Topical Meeting on LOCA, ANL, Chicago, USA, May 25–27th, 2004.
19. A. Stern et al., Investigations of the microstructure and mechanical properties of prior-Beta structure as function of the oxygen content (0.9 wt%) in two zirconium alloys, 15th Int. Symposium on Zirconium in the Nuclear Industry, June 24–28, 2007, Sunriver, Oregon, USA, J. ASTM Int. Vol. 5, No. 4, Paper ID JAI101119, (2008).
20. Perkins · J. Nucl. Mater. · 1977 · 68 · 14810.1016/0022-3115(77)90234-3 ↗
21. Brown · J. Nucl. Mater. · 1982 · 105 · 9310.1016/0022-3115(82)90455-X ↗
22. Wang · Comput. Coupling Phase Diagrams Thermochem. · 2004 · 28 · 28110.1016/j.calphad.2004.09.002 ↗
23. J. Sopoušek, M. Svobodová, Thermodynamic prediction of Zr–Nb–O–H phase diagram sections, in: Proceedings of Solid-Solid Phase Transformations in Inorganic Materials (PTM 2010), Avignon, June 6–11, 2010, p. 652.
24. Okamoto · J. Phase Equilib. · 1992 · 13 · 57710.1007/BF02665776 ↗
25. Négyesi · J. Nucl. Mater. · 2012 · 420 · 31410.1016/j.jnucmat.2011.10.022 ↗
26. M. Negyesi, et al., Contribution to the study of the pseudobinary Zr1Nb–O phase diagram and its application to numerical modeling of the high-temperature steam oxidation of Zr1Nb fuel cladding, Zirconium in the Nuclear Industry STP 1543, 2013,, doi:http://dx.doi.org/10.1520/STP154320120162.10.1520/STP154320120162 ↗
27. V. Vrtilkova, et al., UJP PRAHA a.s. Research Report, in Czech, December 15, 2008.
28. Lee · Z. Metallkd. · 1996 · 87 · 195