Experimental Study of Cr-coated Zirconium Alloy Cladding under Simulated LOCA Conditions

Wang Zhanwei; Yan Jun; Peng Zhenxun; Ren Qisen; Liao Yehong; Li Sigong; Zhao Yahuan

doi:10.13832/j.jnpe.2023.02.0122

Volume 44 Issue 2

Apr. 2023

Turn off MathJax

Article Contents

Article Navigation > Nuclear Power Engineering > 2023 > 44(2): 122-128

Wang Zhanwei, Yan Jun, Peng Zhenxun, Ren Qisen, Liao Yehong, Li Sigong, Zhao Yahuan. Experimental Study of Cr-coated Zirconium Alloy Cladding under Simulated LOCA Conditions[J]. Nuclear Power Engineering, 2023, 44(2): 122-128. doi: 10.13832/j.jnpe.2023.02.0122

Citation:

Wang Zhanwei, Yan Jun, Peng Zhenxun, Ren Qisen, Liao Yehong, Li Sigong, Zhao Yahuan. Experimental Study of Cr-coated Zirconium Alloy Cladding under Simulated LOCA Conditions[J]. Nuclear Power Engineering, 2023, 44(2): 122-128. doi: 10.13832/j.jnpe.2023.02.0122

Citation:

Wang Zhanwei, Yan Jun, Peng Zhenxun, Ren Qisen, Liao Yehong, Li Sigong, Zhao Yahuan. Experimental Study of Cr-coated Zirconium Alloy Cladding under Simulated LOCA Conditions[J]. Nuclear Power Engineering, 2023, 44(2): 122-128. doi: 10.13832/j.jnpe.2023.02.0122

PDF( 32754 KB)

Experimental Study of Cr-coated Zirconium Alloy Cladding under Simulated LOCA Conditions

doi: 10.13832/j.jnpe.2023.02.0122

Department of Nuclear Fuel & Material, China Nuclear Power Technology Research Institute Co., Ltd., Shenzhen, Guangdong, 518026, China

Received Date: 2022-05-17
Rev Recd Date: 2022-12-04
Publish Date: 2023-04-15

Abstract

Abstract

The Fukushima nuclear accident in Japan in 2011 exposed the inherent safety problems of traditional zirconium alloy fuel cladding under LOCA conditions. To investigate the performance of a new Cr-coated zirconium alloy cladding under LOCA conditions, high temperature steam oxidation and quenching experiments under simulated LOCA conditions are carried out for 12~15 μm thick Cr-coated Zr-1Nb alloy cladding tube coated by physical vapor deposition (PVD) process, the oxidation temperature and oxidation time were 1200℃, 1300℃ and 10 min, 20 min, respectively, the quenching was performed around 800℃, then ring compression test was performed for the quenched tube. The results indicated that no spalling was found for Cr coatings under experiment conditions, intense Cr₂O₃ layer which formed on the outer surface of Cr-coated tube retarded the diffusion of O into zirconium substrate, protecting the zirconium alloy from oxidized into ZrO₂ and α-Zr(O) layers, Cr-coated zirconium-alloy cladding remained ductile after quenching. It can be concluded that Cr-coated Zirconium alloy behaves better than traditional Zirconium alloy under the experimental conditions.
- Cr-coated zirconium alloy cladding,
- Loss of coolant accident (LOCA),
- High temperature steam oxidation,
- Quenching,
- Ductile-to-brittle transition

FullText(HTML)

References(18)

References

[1]	YEOM H, LOCKHART C, MARIANI R, et al. Evaluation of steam corrosion and water quenching behavior of zirconium-silicide coated LWR fuel claddings[J]. Journal of Nuclear Materials, 2018, 499: 256-267. doi: 10.1016/j.jnucmat.2017.11.045
[2]	PINT B A, TERRANI K A, YAMAMOTO Y, et al. Material selection for accident tolerant fuel cladding[J]. Metallurgical and Materials Transactions E, 2015, 2(3): 190-196. doi: 10.1007/s40553-015-0056-7
[3]	ZHANG B, GAO P C, XU T, et al. Performance evaluation of accident tolerant fuel under station blackout accident in PWR nuclear power plant by improved ISAA code[J]. Nuclear Engineering and Technology, 2022, 54(7): 2475-2490. doi: 10.1016/j.net.2022.01.024
[4]	YAMAMOTO Y, PINT B A, TERRANI K A, et al. Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors[J]. Journal of Nuclear Materials, 2015, 467: 703-716. doi: 10.1016/j.jnucmat.2015.10.019
[5]	KIM H G, KIM I H, JUNG Y I, et al. Adhesion property and high-temperature oxidation behavior of Cr-coated Zircaloy-4 cladding tube prepared by 3D laser coating[J]. Journal of Nuclear Materials, 2015, 465: 531-539. doi: 10.1016/j.jnucmat.2015.06.030
[6]	MAIER B R, GARCIA-DIAZ B L, HAUCH B, et al. Cold spray deposition of Ti₂AlC coatings for improved nuclear fuel cladding[J]. Journal of Nuclear Materials, 2015, 466: 712-717. doi: 10.1016/j.jnucmat.2015.06.028
[7]	ALAT E, MOTTA A T, COMSTOCK R J, et al. Multilayer (TiN, TiAlN) ceramic coatings for nuclear fuel cladding[J]. Journal of Nuclear Materials, 2016, 478: 236-244. doi: 10.1016/j.jnucmat.2016.05.021
[8]	SHAPOVALOV K, JACOBSEN G M, ALVA L, et al. Strength of SiC_f-SiC_m composite tube under uniaxial and multiaxial loading[J]. Journal of Nuclear Materials, 2018, 500: 280-294. doi: 10.1016/j.jnucmat.2018.01.001
[9]	MAIER B, YEOM H, JOHNSON G, et al. Development of cold spray chromium coatings for improved accident tolerant zirconium-alloy cladding[J]. Journal of Nuclear Materials, 2019, 519: 247-254. doi: 10.1016/j.jnucmat.2019.03.039
[10]	BRACHET J C, IDARRAGA-TRUJILLO I, FLEM M L, et al. Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors[J]. Journal of Nuclear Materials, 2019, 517: 268-285. doi: 10.1016/j.jnucmat.2019.02.018
[11]	BISCHOFF J, DELAFOY C, VAUGLIN C, et al. AREVA NP's enhanced accident-tolerant fuel developments: Focus on Cr-coated M5 cladding[J]. Nuclear Engineering and Technology, 2018, 50(2): 223-228. doi: 10.1016/j.net.2017.12.004
[12]	PARK J H, KIM H G, PARK J Y, et al. High temperature steam-oxidation behavior of arc ion plated Cr coatings for accident tolerant fuel claddings[J]. Surface and Coatings Technology, 2015, 280: 256-259. doi: 10.1016/j.surfcoat.2015.09.022
[13]	BRACHET J C, LE SAUX M, BISCHOFF J, et al. Evaluation of Equivalent Cladding Reacted parameters of Cr-coated claddings oxidized in steam at 1200 C in relation with oxygen diffusion/partitioning and post-quench ductility[J]. Journal of Nuclear Materials, 2020, 533: 152106. doi: 10.1016/j.jnucmat.2020.152106
[14]	YEOM H, MAIER B, JOHNSON G, et al. High temperature oxidation and microstructural evolution of cold spray chromium coatings on Zircaloy-4 in steam environments[J]. Journal of Nuclear Materials, 2019, 526: 151737. doi: 10.1016/j.jnucmat.2019.151737
[15]	PARK D J, KIM H G, JUNG Y I, et al. Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions[J]. Journal of Nuclear Materials, 2016, 482: 75-82. doi: 10.1016/j.jnucmat.2016.10.021
[16]	KIM H G, KIM I H, JUNG Y I, et al. Out-of-pile performance of surface-modified Zr cladding for accident tolerant fuel in LWRs[J]. Journal of Nuclear Materials, 2018, 510: 93-99. doi: 10.1016/j.jnucmat.2018.07.061
[17]	CHUNG H M. Fuel behavior under loss-of-coolant accident situations[J]. Nuclear Engineering and Technology, 2005, 37(4): 327-362.
[18]	RIBIS J, WU A, BRACHET J C, et al. Atomic-scale interface structure of a Cr-coated Zircaloy-4 material[J]. Journal of Materials Science, 2018, 53(14): 9879-9895. doi: 10.1007/s10853-018-2333-1