王栋,钟汝浩,张亚培,郭超,徐浩德,余剑,蓝毅聪,苏光辉,秋穗正,田文喜.磁控溅射铬涂层锆合金包壳高温水蒸气氧化行为[J].表面技术,2023,52(11):258-268.
WANG Dong,ZHONG Ru-hao,ZHANG Ya-pei,GUO Chao,XU Hao-de,YU Jian,LAN Yi-cong,SU Guang-hui,QIU Sui-zheng,TIAN Wen-xi.High-temperature Steam Oxidation Behavior of Magnetron-sputtered Cr-coated Zr Alloy Cladding[J].Surface Technology,2023,52(11):258-268 |
| 磁控溅射铬涂层锆合金包壳高温水蒸气氧化行为 |
| High-temperature Steam Oxidation Behavior of Magnetron-sputtered Cr-coated Zr Alloy Cladding |
| 投稿时间：2022-10-11 修订日期：2023-02-13 |
| DOI：10.16490/j.cnki.issn.1001-3660.2023.11.020 |
| 中文关键词: Zr合金 Cr涂层 事故容错燃料包壳 核反应堆事故 高温水蒸气 氧化动力学 |
| 英文关键词:Zr alloy Cr coating accident tolerant fuel cladding nuclear reactor accident high-temperature steam oxidation kinetics |
| 基金项目:国家重点研发计划(2019YFB1900700) |
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| Author | Institution |
| WANG Dong | School of Nuclear Science and Technology, Xi'an Jiaotong University, Xi'an 710049, China |
| ZHONG Ru-hao | China Nuclear Power Technology Research Institute, Guangdong Shenzhen 518000, China |
| ZHANG Ya-pei | School of Nuclear Science and Technology, Xi'an Jiaotong University, Xi'an 710049, China |
| GUO Chao | China Nuclear Power Technology Research Institute, Guangdong Shenzhen 518000, China |
| XU Hao-de | China Nuclear Power Technology Research Institute, Guangdong Shenzhen 518000, China |
| YU Jian | School of Nuclear Science and Technology, Xi'an Jiaotong University, Xi'an 710049, China |
| LAN Yi-cong | School of Nuclear Science and Technology, Xi'an Jiaotong University, Xi'an 710049, China |
| SU Guang-hui | School of Nuclear Science and Technology, Xi'an Jiaotong University, Xi'an 710049, China |
| QIU Sui-zheng | School of Nuclear Science and Technology, Xi'an Jiaotong University, Xi'an 710049, China |
| TIAN Wen-xi | School of Nuclear Science and Technology, Xi'an Jiaotong University, Xi'an 710049, China |
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| 中文摘要: |
| 目的 研究磁控溅射Cr涂层Zr-1Nb合金包壳在1 100~1 300 ℃水蒸气环境中的氧化行为，为制定核反应堆事故预防和管理提供依据。方法 采用卧式管式炉开展高温氧化试验，通过分析天平测量试样增重，通过扫描电子显微镜观察形貌，通过X射线能谱仪分析元素分布。结果 氧化前Cr涂层结构致密，没有明显缺陷。氧化后包壳表面形成微观的鼓包、褶皱或裂纹。Cr涂层在1 100 ℃和1 200 ℃氧化3 600 s后形成了Cr2O3-Cr-ZrCr2的三层结构。1 200 ℃下，Zr沿Cr晶界扩散到达Cr2O3/Cr界面后将Cr2O3还原，引起局部Cr2O3厚度减小，Cr晶界中的ZrO2则构成了O扩散的短途通道。1 300 ℃氧化1 800 s和3 600 s后，Cr涂层性能退化，生成外侧ZrO2层。在Zr基体氧含量饱和的过程中，ZrO2生长的抛物线常数kp增大。由于包壳内表面氧化使得β-Zr基体达到氧饱和，因此外侧kp迅速进入二次增大阶段，导致外侧ZrO2生长速度明显大于内侧。结论 Cr涂层可以有效提高Zr包壳的抗氧化性能，但经历一定时长高温氧化后将出现性能退化。 |
| 英文摘要: |
| To improve the high-temperature oxidation resistance of Zr alloy cladding under nuclear accident conditions, Cr coating is proposed to be deposited on the cladding surface, which is one of the concepts of accident tolerant fuel (ATF) cladding. In this work, the oxidation behavior of magnetron-sputtered Cr-coated Zr-1Nb alloy cladding in 1 100-1 300 ℃ steam environment was studied. The cladding samples were 9.5 mm in outer diameter, 0.57 mm in thickness and 2 cm in length. Cr coating was deposited on the outer surface of the cladding tube. A horizontal tube furnace was used to carry out the tests. The test conditions included 1 100 ℃/3 600 s, 1 200 ℃/3 600 s, 1 300 ℃/1 800 s and 1 300 ℃/3 600 s. The samples experienced double-sided oxidation during the tests. Mass of the samples was measured by an analytical balance before and after the tests. Surface and cross-section morphologies of the samples were characterized by scanning electron microscopes (SEM). The element distribution was analyzed by energy dispersive spectroscopy (EDS). The as-deposited Cr coating was dense without obvious defects. After oxidation, stress existed in the Cr2O3 layer, which resulted in plastic deformation to form microscopic blisters or folds. If the stress could not be released in time by plastic deformation, micro-cracks appeared on the sample surfaces. Cr2O3 could further react with steam (containing small amount of dissolved O2) to generate volatile products, resulting in the formation of porous surface structure. After oxidation for 3 600 s at 1 100 ℃ and 1 200 ℃, the layered phases of Cr coating from outside to inside were Cr2O3, Cr and ZrCr2, which had a protective effect on the Zr substrate. ZrCr2 was formed by inter-diffusion between the metallic Cr and the Zr substrate. Zr diffused along the grain boundaries in metallic Cr. At 1 200 ℃/ 3 600 s, Zr reached the Cr2O3/Cr interface, and then Cr2O3 was reduced, leading to its local thinning. The diffusion resistance of O in Cr2O3 was reduced in the thinned region, thus resulting in an increased O flux from the Cr2O3/steam interface to the Cr2O3/Cr interface. The Zr combined with O to form ZrO2 precipitates in the grain boundaries of metallic Cr. ZrO2 precipitates acted as short-circuit paths for O to pass through the metallic Cr. Therefore, the amount of O absorbed by Zr substrate increased. After oxidation at 1 300 ℃ for 1 800 s, the degradation of Cr coating occurred and a thick ZrO2 layer grown under the coating. Due to the reduction reaction, the Cr2O3 layer was very thin. Zr was oxidized preferentially to Cr due to its larger oxygen affinity, hence a metallic Cr layer (containing ZrO2 precipitates) was retained. After oxidation for 3 600 s, with the thickening of ZrO2 layer, the metallic Cr was also completely oxidized. During the oxygen saturation of β-Zr and α-Zr(O), the parabolic rate constant for ZrO2 growth (kp) increased. Therefore, the experimental measurement of the thickness of inner ZrO2 layer deviated largely from the calculation by Cathcart-Pawel correlation in the case of 1 300 ℃/3 600 s. Due to the inner-sided oxidation of cladding samples, β-Zr substrate reached oxygen saturation rapidly, thus kp of the outer ZrO2 layer soon entered the secondary increase stage. Therefore, the outer ZrO2 layer had a higher growth rate than the inner one. According to the simulation results, the thickness of the inner and outer ZrO2 layers was reduced compared with the case without coating, indicating an improvement of oxidation resistance by Cr coating. The test conditions in this work are similar to state near the burst regions of cladding tubes during nuclear accidents, thus the results could have a reference value for accident management. |
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