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http://dx.doi.org/10.1016/j.net.2020.05.026

Thermal creep behavior of CZ cladding under biaxial stress state  

Jin, Xin (China Nuclear Power Technology Research Institute Co.,Ltd)
Lin, Yuyu (China Nuclear Power Technology Research Institute Co.,Ltd)
Zhang, Libin (China Nuclear Power Technology Research Institute Co.,Ltd)
Publication Information
Nuclear Engineering and Technology / v.52, no.12, 2020 , pp. 2901-2909 More about this Journal
Abstract
Thermal creep is a key property of zircaloy cladding. CZ developed by CGN is a new zircaloy used as PWR fuel cladding. This research is devoted to investigating the thermal creep behavior of CZ and build the thermal creep model of CZ. Twenty internal pressure creep tests were conducted, and the ranges of temperature and Tresca stress were 320-430 ℃ and 70-300 MPa, respectively. Real-time creep data were analyzed by separating primary creep and steady-state creep. Based on Soderberg model and creep test data, CZ thermal creep model is derived. As a whole, the mean value and the standard deviation of P/M of CZ saturated primary creep strain are very close to these from steady-state creep rate, however, the predictive effect of primary creep is less satisfactory. Four conditions, where there exists large deviation between predicted values and test data, are 320 ℃ and 300 MPa, 350 ℃ and 190 MPa, 380 ℃ and 160 MPa, 380 ℃ and 190 MPa, respectively. As primary creep was much smaller than steady-state creep in long-time operation, the thermal creep model built can be applied to predict the thermal creep behavior of CZ cladding.
Keywords
CZ; Cladding; Thermal creep; Biaxial stress;
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  • Reference
1 Zhilun Xu, Elasticity. Bei Jing, Higher Education Press, 2006, pp. 63-67.
2 Joseph Long, Materials Property Manual of ZIRLO, WEC, 2006.
3 Yong Hwan Jeong, et al., Out-of-pile and in-pile performance of advance zirconium alloys(HANA) for high burnup fuel, J. Nucl. Sci. Technol. 43 (No.9) (2006) 983-997, 2006.
4 P. Bouffioux, et al., Interim dry storage of PWR spent fuel assemblies development of a long term creep law to assess the fuel cladding integrity, in: 8th International Conference on Radioactive Waste Management and Environmental Remediation, 2001.
5 Quecedo, et al., Results of thermal creep test on highly irradiated ZIRLO, Nuclear Engineering and Technology 41 (No.2) (2009) 179-186.
6 Yutaka Matsuo, Thermal creep of zircaloy-4 cladding under internal pressure, J. Nucl. Sci. Technol. 24 (2) (1987) 111-119.
7 Chantal Cappelaere, et al., Thermal creep model for CWSR zircaloy-4 cladding taking into account the annealing of the irradiation hardening, Nucl. Technol. 177 (2011) 257-272.
8 Luscher, Geelhood, Material Property Correlations :Comparisons between FRAPCON-3.4, FRAPTRAN 1.4 and MATPRO, NURGE/CR-7024, 2011.
9 Ito, et al., Evaluation of irradiation effect on spent fuel cladding creep properties, in: Proc. Int. Mtg. LWR Fuel Performance, American Nuclear Society, Orlando, Florida, 2004. September 19-22.
10 Soderberg, Transaction of the ASME 58 (87) (1986) 58-65.
11 Lin Shi, Liutao Chen, Yang Xu, Effect of Final AnnealingTemperature on Axial Creep Property of CZ Alloys, ICNONE26, 2018.
12 Bingye Xu, Plastic Mechanics, Higher Education Press, 1988, pp. 94-95.
13 Yang Xu, Jun Tan, Liutao Chen, Annealing Temperature Effect on Creep Property of CZ Alloy for PWR Modern Fuel Cladding, ICONE25, 2017.
14 zhanying Mu, Creep Mechanics, Xi'an: Publishing House of Xi'an Jiaotong University, 1989, pp. 33-35.
15 Kishore, Effect of hydrogen on the creep behavior of Zr-2.5%Nb alloy at 723 K, J. Nucl. Mater. 385 (3) (2009) 591-594.
16 Bouffioux, Rupa, Impact of Hydrogen on Plasticity and Creep of Unirradiated Zircaloy-4 Cladding Tubes, in: ASTM (Ed.), ASTM Spec. Tech. Publ., 2000, pp. 399-424.
17 Pratik Joshi, et al., Biaxial creep behavior of Nb-modified zircaloys, Nucl. Technol. (2019) 706-716.
18 Limon, Lehmann, A creep rupture criterion for zircaloy-4 fuel cladding under internal pressure, J. Nucl. Mater. (2004) 322-335.
19 Adamson, Garzarolli, Patterson, In-Reactor Creep of zirconium alloys, ZIRAT14 SPECIAL TOPIC REPORT (2009) 43-44.
20 Lucas Bement, Creep of zirconium alloys in nuclear reactors, ASTM STP 815 (1983).
21 Apu Sarkar, et al., Effect of hydrogen on creep behavior of zirconium alloys, JMEPEG (2014) 3649-4656.