Influence of temperature and hydrogen content on stress-induced radial hydride precipitation in Zircaloy-4 cladding
Source abstract
Abstract
Abstract Radial hydride precipitation in stress relieved Zircaloy-4 fuel claddings is studied using a new thermal–mechanical test. Two maximum temperatures for radial hydride precipitation heat treatment are studied, 350 and 450 °C with hydrogen contents ranging between 50 and 600 wppm. The new test provides two main results of interest: the minimum hoop stress required to precipitate radial hydrides and a maximum stress above which, all hydrides precipitate in the radial direction. Based on these two extreme stress conditions, a model is derived to determine the stress level required to obtain a given fraction of radial hydrides after high temperature thermal–mechanical heat treatment. The proposed model is validated using metallographic observation data on pressurized tubes cooled down under constant pressure. Most of the samples with reoriented hydrides are further subjected to a ductility test. Using finite element modeling, the test results are analyzed in terms of crack nucleation within radial hydrides at the outer diameter and crack growth through the thickness of the tubular samples. The combination of test results shows that samples with hydrogen contents of about 100 wppm had the lowest ductility.
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OpenAlex topics: Nuclear Materials and Properties · Nuclear reactor physics and engineering · Fusion materials and technologies
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References (40)
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1. R.P. Marshall, M.R. Louthan, in: Proceedings of the USAEC symposium on Zirconium alloy development, Castlewood, Pleasanton, California, November 12–14, 1962, vol. 1, pp. 14-1.
2. H.M. Chung, in: Proceedings of the 2004 International Meeting on LWR Fuel Performance, Orlando, Florida, September 19–22 2004 (paper 1064).
3. R.S. Daum, S. Madjumdar, Y. Liu, M.C. Billone, in: Proceedings of the 2005 Water Reactor Fuel Performance Meeting, Kyoto, Japan, October 2–6, 2005 (paper 1051).
4. Daum · J. Nucl. Sci. Technol. · 2006 · 43 · 105410.1080/18811248.2006.9711195 ↗
5. Alam · J. ASTM Int. · 2008 · 510.1520/JAI101110 ↗
6. Valance · J. ASTM Int. · 2010 · 810.1520/JAI102957 ↗
7. J. Desquines, D. Drouan, P. March, S. Fourgeaud, C. Getrey, V. Elbaz, M. Philippe, in: Proceedings of the 2005 Water Reactor Fuel Performance Meeting, Charlotte, North Carolina, September 15–19, 2013 (paper 8393).
8. K. Ito, K. Kamimura, Y. Tsukuda, Evaluation of irradiation effect on spent fuel cladding creep properties, in: 2004 International Meeting on LWR Fuel Performance, September 19–22, 2004.
9. Aomi · J. ASTM Int. · 2008 · 510.1520/JAI101262 ↗
10. M. Quecedo, M. Lloret, J.M. Conde, C. Alejano, J.A. Cago, F.J. Fernandez, in: 2008 Water Reactor Fuel Performance Meeting, Seoul, Korea, October 19–23, 2008 (paper 8063).
11. Billone · J. Nucl. Mater. · 2013 · 433 · 43110.1016/j.jnucmat.2012.10.002 ↗
12. V. Busser, J. Desquines, C. Duriez, in: Proceedings of Topfuel 2009, Paris, France, September 6–10, 2009 (paper 2104).
13. ASTM, ASTM G38-73 Standard practice for making and using C-Ring stress-corrosion test specimens, 1995.
14. ASTM, ASTM C1323-96 Standard test method for ultimate strength of advanced ceramics with diametrally compressed C-Ring specimens at ambient temperature, 1996.
15. http://www-cast3m.cea.fr/, 2014.
16. Martin-Rengel · J. Nucl. Mater. · 2013 · 436 · 12310.1016/j.jnucmat.2013.01.311 ↗
17. Kearns · J. Nucl. Mater. · 1966 · 20 · 24110.1016/0022-3115(66)90036-5 ↗
18. Une · J. Nucl. Mater. · 2003 · 322 · 6610.1016/S0022-3115(03)00320-9 ↗
19. Kearns · J. Nucl. Mater. · 1967 · 22 · 29210.1016/0022-3115(67)90047-5 ↗
20. Pan · J. Nucl. Mater. · 1996 · 228 · 22710.1016/S0022-3115(95)00217-0 ↗
21. Puls · J. Nucl. Mater. · 2009 · 393 · 35010.1016/j.jnucmat.2009.06.022 ↗
22. Puls · J. Nucl. Mater. · 2010 · 399 · 22710.1016/j.jnucmat.2010.01.026 ↗
23. Hardie · J. Nucl. Mater. · 1975 · 55 · 110.1016/0022-3115(75)90132-4 ↗
24. C.D. Wiliams, M.O. Marlowe, R.B. Adamson, S.B. Wisner, R.A. Rand, S. Armijo, in: Zirconium in the Nuclear Industry: Eleventh International Symposium, ASTM STP 1295, 1996, pp. 676–694.10.1520/STP16196S ↗
25. Goll · J. Nucl. Mater. · 2001 · 289 · 24710.1016/S0022-3115(01)00438-X ↗
26. C. Cappelaere, R. Limon, D. Gilbon, T. Bredel, O. Rabouille, P. Bouffioux, J.-P. Mardon, in: Zirconium in the Nuclear Industry: Thirteenth International Symposium, ASTM STP 1423, 2002, pp. 720–739.10.1520/STP11413S ↗
27. M. Billone, T.A. Bursteva, L. Liu, Used Fuel Disposition Campaign Baseline Studies for Ring Compression Testing of High Burnup Fuel Cladding, Argonne National Laboratory report, ANL-12-58, 2012.10.2172/1121037 ↗
28. P. Bouffioux, Réorientation des hydrures – Synthèse des résultats acquis pour les gaines en Zircaloy-4 AFA-2G, EDF Report, HT25-C2002-191/PB, 2002.
29. A. Racine, M. Bornert, C. Sainte-Catherine, C. Cappelaere, in: 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT 18) – Beijing, China, August 7–12, 2005.
30. Ells · J. Nucl. Mater. · 1968 · 35 · 30610.1016/0022-3115(70)90214-X ↗
31. M.P. Puls, Hydrogen-induced delayed cracking – 2. Effect of stress on nucleation growth and coarsening of zirconium hydride precipitates, AECL Report 8381, 1985.
32. Puls · 1986 · 426
33. Qin · Acta Mater. · 2011 · 59 · 701010.1016/j.actamat.2011.07.054 ↗
34. Tuinet · Intermetallics · 2012 · 20 · 2410.1016/j.intermet.2011.08.005 ↗
35. MacEwen · Metall. Trans. A · 1990 · 21A · 108310.1007/BF02656529 ↗
36. Bai · Metall. Mater. Trans. A · 1994 · A25 · 119910.1007/BF02652294 ↗
37. Massih · Comput. Mater. Sci. · 2009 · 4 · 109110.1016/j.commatsci.2009.05.025 ↗
38. Chu · J. Nucl. Mater. · 2008 · 373 · 31910.1016/j.jnucmat.2007.06.012 ↗
39. Sakamoto · J. Nucl. Sci. Technol. · 2006 · 43 · 113610.1080/18811248.2006.9711205 ↗
40. Shi · J. Nucl. Mater. · 1999 · 275 · 31210.1016/S0022-3115(99)00132-4 ↗