泡核沸腾对高温富氧水中Zr合金腐蚀行为的影响

doi:10.3901/JME.2024.24.199

摘要/Abstract

摘要： 为了揭示空泡份额对高温富氧水中Zr合金表面水垢沉积和腐蚀行为的影响规律，采用高温高压动水回路试验，对Zr-4和SZA-4合金包壳管开展42 d腐蚀行为的对比评价研究。通过扫描电镜(SEM)和透射电镜(TEM)对Zr合金氧化膜及水垢的显微组织进行了观察，采用氢含量分析仪对吸氢量进行测定。结果表明，随着沿包壳管轴向温度和空泡份额的提高，两种Zr合金氧化膜厚度均表现出增加趋势，氧化膜/基体(O/M)界面的起伏程度加剧，同时吸氢量随之增加。腐蚀42 d时，两种Zr合金均处于腐蚀转折前的早期阶段，氧化膜保持良好的致密性及保护性，外层为等轴晶区，内层为柱状晶区，分布着较多尚未被完全氧化的第二相析出粒子(SPPs)，同时O/M界面存在明显的过渡层。包壳管表面水垢主要由不同晶粒尺寸的Fe的氧化物组成，可分为“疏松区”和“致密区”，其中疏松区的晶粒细小，存在大量孔洞。含Nb的SZA-4合金的氧化膜厚度大于无Nb的Zr-4合金，高温富氧水中的耐腐蚀性能较差，证实了Nb元素增加富氧环境中Zr合金腐蚀速率的事实，而这与常规压水堆环境中的腐蚀规律相反。

关键词: Zr合金, 腐蚀, 氧化膜, 溶解氧, 空泡份额, 水垢

Abstract: In order to understand the effect of void fraction on CRUD deposition and corrosion behavior of zirconium alloy in high temperature water with dissolved oxygen, a 42-days corrosion test was carried out on Zr-4 and SZA-4 alloy cladding tubes in high temperature and high pressure water loop system. The microstructures of oxide and CRUD were analyzed by scanning electron microscope (SEM) and transmission electron microscope (TEM). The hydrogen absorbed by Zr metal substrate was examined by a hydrogen analyzer. The results show that, with the increase of temperature and void fraction along the cladding tube, the oxide film thickness of the two Zr alloys generally grow, the oxide-metal interface become uneven and hydrogen absorption increase accordingly. After corrosion for 42 days, both Zr-4 and SZA-4 alloys are still in the early corrosion stage before transition. The oxide consists of outer equiaxed grains and inner columnar grains, which is dense and totally protectable. There are several second phase particles (SPPs) without full oxidization distributing across the whole oxide, and a transition layer was observed at the oxide and α-Zr matrix (O/M) interface. The CRUD is mainly composed of iron oxide crystals with different grain sizes, and can be divided into loose and dense region, and there are a lot of holes in the loose region with small grain size. The oxide thickness for Nb-containing SZA-4 alloy is greater than that of Zr-4 alloy without Nb, indicating the inferior corrosion resistance of SZA-4 alloy in oxygen-enriched high temperature water. The phenomena confirm that Nb element increases the corrosion rate of Zr alloy when some dissolved oxygen is present in water, which is contrary to the corrosion behavior in conventional pressurized water reactor environment.

Key words: zirconium alloys, corrosion, oxide, dissolved oxygen, void fraction, CRUD

中图分类号:

TL341

曾奇锋, 刘妍洁, 张乐福, 司胜义, 李聪, 卢俊强, 刘庆冬. 泡核沸腾对高温富氧水中Zr合金腐蚀行为的影响[J]. 机械工程学报, 2024, 60(24): 199-210.

ZENG Qifeng, LIU Yanjie, ZHANG Lefu, SI Shengyi, LI Cong, LU Junqiang, LIU Qingdong. Effect of Nucleate Boiling on Corrosion Behavior of Zirconium Alloy in Oxygen-enriched High Temperature Water[J]. Journal of Mechanical Engineering, 2024, 60(24): 199-210.

参考文献

[1] HU J，SETIADINATA B，AARHOLT T，et al. Understanding corosion and hydrogen pickup of zirconium fuel cladding alloys：the role of oxide microstructure，porosity，suboxides，and second-phase particles[C]//Zirconium in the Nuclear Industry：18th International Symposium. Eagan，MN：ASTM International，2018：93-126.
[2] JIANG G，XU D，YANG W，et al. High-temperature corrosion of Zr–Nb alloy for nuclear structural materials[J]. Progress in Nuclear Energy，2022，154：104490.
[3] JEONG Y H，KIM H G，KIM D J，et al. Influence of Nb concentration in the α-matrix on the corrosion behavior of Zr–xNb binary alloys[J]. Journal of Nuclear Materials，2003，323(1)：72-80.
[4] 杨忠波，赵文金. 锆合金耐腐蚀性能及氧化特性概述[J]. 材料导报，2010，24(17)：120-125. YANG Zhongbo，ZHAO Wenjin. Review of corrosion and oxide characterization for Zr alloys[J]. Materials Report，2010，24(9)：120-125.
[5] 周邦新. 锆合金中的疖状腐蚀问题[J]. 核科学与工程，1993，13(1)：51-58. ZHOU Bangxin. The problems of nodular corrosion in Zircaloy[J]. Chinese Journal of Nuclear Science and Engineering，1993，13(1)：51-58.
[6] 周邦新，姚美意，李强，等. Zr-Sn-Nb合金耐疖状腐蚀性能的研究[J]. 稀有金属材料与工程，2007，36(8)：1317-1321. ZHOU Bangxin，YAO Meiyi，LI Qiang，et al. Nodular corrosion resistance of Zr-Sn-Nb alloy[J]. Rare Metal Materials and Engineering，2007，36(8)：1317-1321.
[7] NOUDURU S K，MANDAPAKA K K，ROYCHOWDHURY S，et al. Nodular corrosion of zirconium alloys in gaseous environment containing different contaminants[J]. Journal of Nuclear Materials，2021，545：152640.
[8] 岳雅楠，陈寰，余施佳，等. 锆合金Cr涂层包壳管在环向压缩作用下的裂纹行为[J].中国表面工程，2022，35(4)：1-9.YUE Yanan，CHEN Huan，YU Shijia，et al. Crack behavior of Cr coating zircaloy cladding tubes under ring compression[J]. China Surface Engineering，2022，35(4)：1-9.
[9] CHEN C M，ZHOU B X，GOU S Q，et al. Effect of surface state on nodular corrosion resistance of Zircaloy-4 alloy[J]. Acta Metallurgica Sinica(English Letters)，2015，28(2)：254-260.
[10] KUMAR M K，AGGARWAL S，KAIN V，et al. Effect of dissolved oxygen on oxidation and hydrogen pick up behaviour-Zircaloy vs Zr–Nb alloys[J]. Nuclear Engineering and Design，2010，240(5)：985-994.
[11] BRADHURST D H，SHIRVINGTON P J，HEUER P M. The effects of radiation and oxygen on the aqueous oxidation of zirconium and its alloys at 290℃[J]. Journal of Nuclear Materials，1973，46(1)：53-76.
[12] 刘庆冬，张浩，曾奇锋，等. SZA-4和ZIRLO锆合金在360℃含氧水环境中的腐蚀行为[J]. 机械工程学报，2019，55(8)：88-96. LIU Qingdong，ZHANG Hao，ZENG Qifeng，et al. Pre-transition corrosion behavior of SZA-4 and ZIRLO alloys in dissolved oxygen aqueous condition at 360℃[J]. Journal of Mechanical Engineering，2019，55(8)：88-96.
[13] LAI P，LU J，ZHANG H，et al. The corrosion behavior of M5(Zr-1Nb-0.12O) alloy in 360℃ water with dissolved oxygen[J]. Journal of Nuclear Materials，2020，532：152079.
[14] XU S，BAI Y，HUANG J，et al. Uniform corrosion behavior of Zr-1Sn-0.35Fe-0.15Cr -xNb alloys in 400℃ super-heated steam with oxygen[J]. Corrosion Science，2023，214：111004.
[15] XU S，HUANG J，PEI W，et al. Effect of oxygen content in 400 ℃ super-heated steam on the corrosion resistance of Zr-xSn-0.35Fe-0.15Cr -0.15Nb alloys[J]. Corrosion Science，2022，198：110135.
[16] SUN R，XU S，YAO M，et al. Effect of dissolved oxygen on corrosion behavior of Zr-0.85Sn-0.16Nb-0.37Fe-0.18Cr alloy in 500℃ and 10.3 MPa super-heated steam[J]. Transactions of Nonferrous Metals Society of China，2020，30(3)：701-709.
[17] 韦天国，林建康，龙冲生，等. 蒸汽中的溶解氧对锆合金腐蚀行为的影响[J]. 金属学报，2016，52(2)：209-216. WEI Tianguo，LIN Jiankang，LONG Chongsheng，et al. Effect of dissolved oxygen in steam on the corrosion behaviors of zirconium alloys[J]. Acta Metallurgica Sinica，2016，52(2)：209-216.
[18] BILLOT P，BESLU P，GIORDANO A，et al. Development of a mechanistic model to assess the external corrosion of the zircaloy claddings in PWRs[C]//Zirconium in the Nuclear Industry：Eighth International Symposium，Baltimore，MD：ASTM International，1989：165-184.
[19] BOUINEAU V，AMBARD A，BÉNIER G，et al. A new model to predict the oxidation kinetics of zirconium alloys in a pressurized water reactor[C]//Zirconium in the Nuclear Industry：Fifteenth International Symposium，Baltimore，MD：ASTM International，2009：405- 429.
[20] 曾奇锋，朱丽兵，袁改焕，等. CAP1400燃料组件用新锆合金研究[J]. 核技术，2017，40(3)：61-67. ZENG Qifeng，ZHU Libing，YUAN Gaihuan，et al. Study on new zirconium alloys for CAP1400 fuel assembly[J]. Nuclear Techniques，2017，40(3)：61-67.
[21] WEI T，DAI X，ZHAO Y，et al. ZrO phase embedded in the oxide of Zr-Sn-Nb-Fe-Cr alloy after corrosion[J]. Journal of Nuclear Materials，2022，571：153992.
[22] HU J，GARNER A，NI N，et al. Identifying suboxide grains at the metal-oxide interface of a corroded Zr- 1.0%Nb alloy using (S)TEM，transmission-EBSD and EELS[J]. Micron，2015，69：35-42.
[23] LIAO J，YANG Z，QIU S，et al. The correlation between tetragonal phase and the undulated metal/oxide interface in the oxide films of zirconium alloys [J]. Journal of Nuclear Materials，2019，524：101-110.
[24] PROFF C，ABOLHASSANI S，LEMAIGNAN C. Oxidation behaviour of zirconium alloys and their precipitates-a mechanistic study[J]. Journal of Nuclear Materials，2013，432：222-238.
[25] COX B. Some thoughts on the mechanisms of in-reactor corrosion of zirconium alloys[J]. Journal of Nuclear Materials，2005，336：331-368.
[26] CHEN L，ZENG Q，LI J，et al. Effect of microstructure on corrosion behavior of a Zr-Sn-Nb-Fe-Cu-O alloy[J]. Materials and Design，2016，92：888-896.
[27] HUANG J，YAO M，GAO C，et al. The influence of second phase particles on the crack formation in oxide films formed on zirconium alloys[J]. Corrosion Science，2015，99：172-177.
[28] NI N，HUDSON D，WEI J，et al. How the crystallography and nanoscale chemistry of the metal/oxide interface develops during the aqueous oxidation of zirconium cladding alloys[J]. Acta Materialia，2012，60：7132-7149.
[29] 刘建章. 核结构材料[M]. 北京：化学工业出版社，2007. LIU Jianzhang. Nuclear structural materials[M]. Beijing：Chemical Industry Press，2007.
[30] GARNER A，HU J，HARTE A，et al. The effect of Sn concentration on oxide texture and microstructure formation in zirconium alloys[J]. Acta Materialia，2015，99：259-272.
[31] WEI J，FRANKEL P，POLATIDIS E，et al. The effect of Sn on autoclave corrosion performance and corrosion mechanisms in Zr–Sn–Nb alloys[J]. Acta Materialia，2013，61(11)：4200-4214.
[32] MATSUKAWA Y，KITAYAMA S，MURAKAMI K，et al. Reassessment of oxidation-induced amorphization and dissolution of Nb precipitates in Zr-Nb nuclear fuel cladding tubes[J]. Acta Materialia，2017，127：153-164.