Experimental Study on Microstructure and Wetting Properties of N36 Zirconium Alloy Oxide Layer
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摘要: 研究了国产N36锆合金包壳在600、700℃和800℃常压下形成的氧化层微观形貌和表面润湿特性。对N36锆合金样件进行氧化，并测量了氧化层厚度和表面接触角。对样件表面进行扫描电子显微镜（SEM）观测获得样件的表面微观形貌，利用能谱仪（EDS）对样件表面进行局部扫描获得了成分元素种类和含量分布，分析了氧化温度和氧化时间对于N36锆合金表面润湿性的影响规律。结果表明，氧化后的样件表面润湿性增强，氧化层表面裂纹的尺寸、深度、内部结构都会影响表面润湿性。随着氧化温度升高，裂纹尺寸有增加的趋势。在同一氧化温度下，随着氧化时间的增长，样件表面裂纹的尺寸和数量都有增加的趋势。本文研究有助于深入了解N36锆合金包壳材料表面氧化的微观特性。Abstract: The microstructure and surface wetting properties of domestic N36 zirconium alloy cladding under atmospheric pressure at 600℃, 700℃ and 800℃ are studied. The N36 sample is oxidized, and the oxide layer thickness and surface contact angle are measured. The surface microstructure of the sample is observed by scanning electron microscope (SEM). The type and content distribution of elements are obtained by local scanning of the sample surface by energy dispersive spectrometer (EDS). The effects of oxidation temperature and oxidation time on the surface wetting properties of N36 zirconium alloy are analyzed. The results show that the surface wetting properties of the oxidized sample is enhanced, and the size, depth and internal structure of the surface crack of the oxide layer will affect the surface wetting properties. With the increase of oxidation temperature, the crack size tends to increase. At the same oxidation temperature, the size and number of surface cracks increase with the increase of oxidation time. The study of this paper is helpful to further understand the microscopic characteristics of surface oxidation of N36 zirconium alloy cladding materials.
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Key words:
- N36 zirconium alloy /
- Oxide layer /
- Surface microstructure /
- Surface wetting properties
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图 1 高温蒸汽氧化实验平台示意图
Figure 1. Schematic Diagram of High-temperature Steam Oxidation Experimental Platform
图 2 圆管静态接触角θ示意图
Figure 2. Schematic Diagram of Static Contact Angle θ on Circular Tube
图 3 600℃和700℃下氧化样件的氧化层厚度、接触角与表面Zr-O比
Figure 3. Oxidation Layer Thickness, Contact Angle and Surface Zr-O Ratio of Oxidation Samples at 600℃ and 700℃
图 4 600℃下氧化1 h样件表面EDS分析结果
Figure 4. EDS Analysis Results of Sample Surface Oxidized for 1 h at 600 ℃
图 5 600℃下氧化1~4 h的样件表面SEM图
Figure 5. SEM Images of Sample Surface Oxidized for 1~4 h at 600℃
图 6 600℃下氧化1~4 h的样件截面SEM图
Figure 6. SEM Images of Sample Section Oxidized for 1~4 h at 600℃
图 7 700℃下氧化1~4 h的样件表面SEM图
Figure 7. SEM Images of Sample Surface Oxidized for 1~4 h at 700℃
图 8 700℃下氧化1~4 h的样件截面SEM图
Figure 8. SEM Images of Sample Section Oxidized for 1~4 h at 700℃
图 9 800℃下氧化1 h的样件表面SEM图和截面SEM图
Figure 9. SEM Images of Sample Surface and Section Oxidized for 1 h at 800℃
表 1 600℃下氧化1 h样件表面元素分布
Table 1. Element Distribution on Sample Surface Oxidized for 1 h at 600℃
元素 特征X射
线类型重量百
分比/%重量百
分比标
准偏差原子百
分比/%氧化物 氧化物
百分含
量/%氧化物百
分含量标
准偏差O K线系 25.85 0.33 66.21 Zr L线系 71.71 0.43 32.22 ZrO2 96.87 0.58 Sn L线系 0.56 0.20 0.20 SnO2 0.72 0.26 Fe L线系 1.88 0.36 1.38 FeO 2.42 0.46 总量 100.00 100.00 100.00 
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[1] 高巍,张娴,王正品,等. M5和Zirlo合金高温水蒸气氧化行为研究[J]. 西安工业大学学报,2016, 36(6): 473-480. doi: 10.16185/j.jxatu.edu.cn.2016.06.008 [2] 卓洪,邱绍宇,赵文金,等. 军民结合助推N36锆合金实现工程化应用[J]. 中国军转民,2019(4): 73-76. doi: 10.3969/j.issn.1008-5874.2019.04.025 [3] SUN C, YANG Z B, WU Z P. Study on corrosion resistance of N36 zirconium alloy in LiOH aqueous solution[J]. World Journal of Nuclear Science and Technology, 2018, 8(2): 30-37. doi: 10.4236/wjnst.2018.82004 [4] 苗一非,焦拥军,张坤,等. N36锆合金包壳堆内腐蚀模型研究[J]. 原子能科学技术,2018, 52(2): 290-294. [5] 程竹青,沈剑韵,陈波全,等. N36锆合金相图计算初步研究[J]. 核动力工程,2020, 41(S1): 147-152. [6] 惠泊宁,渠静雯,李帆,等. 表层渗碳对N36锆合金管坯耐腐蚀性能的影响[J]. 有色金属加工,2019, 48(4): 11-13,57. doi: 10.3969/j.issn.1671-6795.2019.04.004 [7] 苗一非,焦拥军,张坤,等. N36锆合金包壳辐照生长经验模型研究[J]. 原子能科学技术,2019, 53(2): 277-281. doi: 10.7538/yzk.2018.youxian.0303 [8] PAWEL R E, CATHCART J V, CAMPBELL J J, et al. Zirconium metal-water oxidation kinetics. V. Oxidation of Zircaloy in high pressure steam. [PWR]: ORNL/NUREG-31[R]. Tenn: Oak Ridge National Lab., 1977. [9] 李冬,许巍,刘晓晶,等. RELAP5程序基于敏感性分析的再淹没模型改进[J]. 核动力工程,2015, 36(S2): 151-156. [10] KANDLIKAR S G. A theoretical model to predict pool boiling CHF incorporating effects of contact angle and orientation[J]. Journal of Heat Transfer, 2001, 123(6): 1071-1079. doi: 10.1115/1.1409265 [11] CHEN M J, WU D, CHEN D Q, et al. Experimental investigation on the movement of triple-phase contact line during a droplet impacting on horizontal and inclined surface[J]. Chemical Engineering Science, 2020, 226: 115864. doi: 10.1016/j.ces.2020.115864 [12] MAKI H. Heat transfer characteristics of zircaloy-2 oxide film[J]. Journal of Nuclear Science and Technology, 1973, 10(3): 170-175. doi: 10.1080/18811248.1973.9735399 [13] GÖHR H, SCHALLER J, SCHILLER C A. Impedance studies of the oxide layer on zircaloy after previous oxidation in water vapour at 400℃[J]. Electrochimica Acta, 1993, 38(14): 1961-1964. doi: 10.1016/0013-4686(93)80323-R [14] URBANIC V F, HEIDRICK T R. High-temperature oxidation of zircaloy-2 and zircaloy-4 in steam[J]. Journal of Nuclear Materials, 1978, 75(2): 251-261. doi: 10.1016/0022-3115(78)90006-5 [15] NAGASE F, OTOMO T, UETSUKA H. oxidation kinetics of low-Sn zircaloy-4 at the temperature range from 773 to 1, 573 K[J]. Journal of Nuclear Science and Technology, 2003, 40(4): 213-219. doi: 10.1080/18811248.2003.9715351 [16] LEISTIKOW S, SCHANZ G. Oxidation kinetics and related phenomena of zircaloy-4 fuel cladding exposed to high temperature steam and hydrogen-steam mixtures under PWR accident conditions[J]. Nuclear Engineering and Design, 1987, 103(1): 65-84. doi: 10.1016/0029-5493(87)90286-X [17] 张喜燕. 熔化情况下Zr-4合金包壳管氧化膜厚度变化的计算[J]. 核动力工程,1991, 12(2): 83-87. [18] YEOM H, JO H, JOHNSON G, et al. Transient pool boiling heat transfer of oxidized and roughened Zircaloy-4 surfaces during water quenching[J]. International Journal of Heat and Mass Transfer, 2018, 120: 435-446. doi: 10.1016/j.ijheatmasstransfer.2017.12.060 [19] KANG J Y, KIM T K, LEE G C, et al. Quenching of candidate materials for accident tolerant fuel-cladding in LWRs[J]. Annals of Nuclear Energy, 2018, 112: 794-807. doi: 10.1016/j.anucene.2017.11.007 [20] MASSIH A R. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions: SKI-R-07-14[R]. Stockholm: Swedish Nuclear Power Inspectorate, 2007. [21] 安欣林,王晓莉. 纳米二氧化锆研究进展−Ⅰ性质[J]. 内蒙古石油化工,2007(5): 28-30. doi: 10.3969/j.issn.1006-7981.2007.05.012 -
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