Modeling of Zircaloy cladding primary creep during load drop and reversal
Source abstract
Abstract
Modeling fuel behavior requires an accurate description of the cladding stress response for both operational and safety considerations. The transient creep response of Zirconium alloys is commonly modeled using a strain hardening rule which is known to hold in cases with monotonously increasing stresses. However, the strain hardening rule is experimentally known to fail in scenarios such as load drop or reversal. In this paper we derive a simple and easily implementable set of rules for primary creep based on experimental results which contradict the strain hardening rule. The primary creep predicted by these rules is compared with data from published thermal creep experiments and Halden in-pile creep experiment IFA-585. The model thus created is shown to perform well in describing both transient stress scenarios with monotonously increasing stress and scenarios involving load drops and reversals.
Source: openaire · Quality: candidate unverified · Source record ↗
Research context
OpenAlex topics: Nuclear Materials and Properties · Nuclear reactor physics and engineering · Nuclear and radioactivity studies
SciVX discovery tags are terminology-based. OpenAlex topics are source-provided classifications.
References (29)
Show source-provided references
1. Wiesenack · Journal of Nuclear Science and Technology · 2006 · 43 · 103710.1080/18811248.2006.9711193 ↗
2. Lucas · Nuclear Technology · 1981 · 53 · 4610.13182/NT81-A17055 ↗
3. Lee · Journal of Nuclear Materials · 1980 · 88 · 10410.1016/0022-3115(80)90390-6 ↗
4. L. Jernkvist, in: Transactions of the 15th International Conference on Structural Mechanics in Reactor Technology (SMiRT-15) C04/3, 1999, pp. 477–484.
5. K. Murty, K. Yoon, in: Transactions of the 5th International Conference on Structural Mechanics in Reactor Technology (SMiRT-5) C3/6, 1979.
6. Murty · JOM · 1999 · 51 · 3210.1007/s11837-999-0184-6 ↗
7. T. Ikonen, V. Tulkki, Nuclear Engineering and Design, submitted for publication.
8. Rossiter · Nuclear Engineering and Technology · 2011 · 43 · 48910.5516/NET.2011.43.6.489 ↗
9. M. McGrath, HWR-471, 1996.
10. Y. Matsuo, ASTM STP 1023, 1989, pp. 678–691.10.1520/STP18893S ↗
11. D. Franklin, G. Lucas, A. Bement, ASTM STP 815, 1983.
12. Matsuo · Journal of Nuclear Science and Technology · 1987 · 24 · 11110.1080/18811248.1987.9735783 ↗
13. F. Garzarolli, H. Stehle, E. Steinberg, ASTM STP 1295, 1996, pp. 12–32.10.1520/STP16165S ↗
14. Kim · Journal of Nuclear Materials · 1997 · 250 · 16410.1016/S0022-3115(97)00261-4 ↗
15. A. Soniak, N. L’Hullier, J.-P. Mardon, V. Rebeyrolle, P. Bouffioux, C. Bernaudat, ASTM STP 1423, 2002, pp. 837–862.10.1520/STP11419S ↗
16. K. Ito, K. Kamimura, Y. Tsukuda, in: Proceedings of the 2004 International Meeting on LWR Fuel Performance, 2004, pp. 440–451.
17. M. Griffiths, N. Christodoulou, S. Donohue, ASTM STP 1467, 2005, p. 686.10.1520/STP37531S ↗
18. Moon · Journal of Nuclear Materials · 2006 · 353 · 17710.1016/j.jnucmat.2006.01.023 ↗
19. Holt · Journal of Nuclear Materials · 2008 · 372 · 18210.1016/j.jnucmat.2007.02.017 ↗
20. Wang · Journal of Nuclear Materials · 2013 · 433 · 18810.1016/j.jnucmat.2012.08.049 ↗
21. M. Limback, T. Andersson, ASTM STP 1295, 1996, pp. 448–468.10.1520/STP16185S ↗
22. Geelhood · Nuclear Engineering and Technology · 2011 · 43 · 50910.5516/NET.2011.43.6.509 ↗
23. K. Geelhood, W. Luscher, C. Beyer, NUREG/CR-7022, vol. 1, 2011.
24. W. Luscher, K. Geelhood, NUREG/CR-7024, 2011.
25. J. Foster, M. McGrath, in: Proceedings of the 2007 LWR Fuel Performance Meeting, 2007.
26. Banks · Advances in Applied Mathematics and Mechanics · 2011 · 3 · 110.4208/aamm.10-m1030 ↗
27. IFPE database NEA IFPE/IFA-585 experimental data, Last modified 13-MAR-2008.
28. A. Donaldson, HWR-413, 1994.
29. M. McGrath, in: Proceedings of the 2000 International Topical Meeting on LWR Fuel Performance, 2000.