目的 流致振动诱发的微动磨损是导致锆合金包壳失效的主要因素,揭示多维度耦合模式下包壳的磨损机理、延长其服役寿命是安全发展核能亟须解决的重点问题。研究核用Zr -Sn-Nb合金包壳在不同冲击能量下的冲-切磨损行为,以及预氧化处理对其耐磨性能的提升机制。方法 采用特制高压釜对锆合金包壳进行预氧化处理,在其表面制备ZrO2层。利用特制的冲击-切向磨损试验设备进行冲切磨损试验,获取包壳在不同冲击能量下的冲切磨损规律,揭示预氧化处理技术对其抗磨损性能的提升机制。结果 经预氧化处理后,包壳硬度为233.5HV0.2,相较于基体(191.0HV0.2)提高了22.3%,表面粗糙度为0.306 μm,相较于基体(0.458 μm)降低了33.2%。当EI=3.60 mJ时,预氧化包壳的冲击回弹速度、能量和冲击力略高于基体,而能量吸收率、切向摩擦力、摩擦因数和磨损体积相较于基体分别由39.1%、53.9 N、0.59、8.45×105 μm3降至37.7%、44.90 N、0.39、8.01×105 μm3。结论 随着冲击能量的增加,基体的磨损机制由磨粒磨损和轻微疲劳磨损转变为以分层为主;预氧化涂层有效改善了包壳的表面质量,硬度升高,粗糙度降低,分别降低了冲切过程中的能量吸收率和切向摩擦因数,使其磨损机理转变为以轻微磨粒磨损和疲劳磨损为主,有效提高了其抗冲切磨损性能。
Abstract
The zirconium alloy cladding is the core component of the reactor in the nuclear power plant. During the service process, the cladding is affected by the axial and transverse flow formed by the high temperature, high pressure, and high-velocity water, which is easy to undergo impact-sliding coupling wear along the axial and radial directions with the positioning grid, resulting in the early failure of the cladding and the leakage of fission products. The work aims to investigate the impact-sliding wear behavior of fuel cladding under different impact energies and the mechanism of pre-oxidation treatment to improve its wear resistance. Firstly, the fuel cladding was pre-oxidized by a special autoclave. Before the wear test, the hardness, surface roughness, and phase composition of the pre-oxidized layer were characterized and tested by a Vickers hardness meter, white light interferometer, and X-ray diffractometer, and the surface morphology and thickness of the oxidative layer were measured through optical microscope, scanning electron microscope, and energy dispersive spectrometer. Finally, the special impact-sliding wear test equipment was used to carry out the impact-sliding wear test on the cladding before and after the pre-oxidation treatment under three different impact energies, to obtain the impact wear rule of the cladding under different impact energies and reveal the mechanism of improving the anti-wear performance of the pre-oxidation treatment technology. A parallel line contact form of tube/plane was adopted in the test and the test parameters were as follows: the sliding velocity was 90 mm/s, the impact velocity was 60, 90, and 120 mm/s, respectively and the number of cycles was 1×104. After the test, a white light interferometer was used to test the wear scars, and the 3D morphologies of wear scars, cross-section profiles, wear area, and wear volume were obtained. The morphologies of wear scars were obtained by optical microscope and scanning electron microscope. After pre-oxidation treatment, the hardness of the cladding was about 233.5HV0.2, which was 22.3% higher than that of the substrate (191.0HV0.2), and the surface roughness was 0.306 μm, which was 33.2% lower than that of the substrate (0.458 μm). The results showed that the impact force, friction, friction coefficient, absorbed energy, and wear volume of the cladding increased with the increase of impact energy whether before or after the pre-oxidation treatment, and the energy absorption rate increased firstly and then decreased. When the impact energy EI was 3.60 mJ, the impact rebound speed, energy, and impact force of the pre-oxidized cladding were slightly higher than that of the substrate. The energy absorption rate, tangential friction force, friction coefficient, and wear volume decreased from 39.1%, 53.9 N, 0.59, and 8.45×105 μm3 to 37.7%, 44.90 N, 0.39, and 8.01×105 μm3, respectively. The pre-oxidation treatment technology was helpful to reduce the wear degree of the cladding. Besides, with the increase of impact energy, the wear mechanism of the uncoated cladding changed from abrasive wear and mild fatigue wear to delamination. The pre-oxidation treatment can effectively increase the rebound speed of the cladding, reduce the friction force, decrease the friction coefficient and energy absorption rate, and finally reduce the wear volume, change the wear mechanism into slight abrasive wear and fatigue wear, and effectively improve the impact-sliding wear resistance.
关键词
燃料棒 /
锆合金 /
预氧化 /
冲-切磨损 /
耐磨性 /
磨损机理
Key words
fuel rod /
zirconium alloy /
pre-oxidation /
impact-sliding wear /
wear resistance /
wear mechanism
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参考文献
[1] WANG J, LEI Y J, LI Z Y, et al.Effect of Contact Misalignment on Fretting Wear Behavior between Fuel Cladding and Zr-4 Grid[J]. Tribology International, 2023, 181: 108299.
[2] LEE Y H, KIM H K.Fretting Wear Behavior of a Nuclear Fuel Rod under a Simulated Primary Coolant Condition[J]. Wear, 2013, 301(1/2): 569-574.
[3] MOTTA A T, COUET A, COMSTOCK R J.Corrosion of Zirconium Alloys Used for Nuclear Fuel Cladding[J]. Annual Review of Materials Research, 2015, 45: 311-343.
[4] KIM H K, KIM S J, YOON K H, et al.Fretting Wear of Laterally Supported Tube[J]. Wear, 2001, 250(1/2/3/4/5/6/ 7/8/9/10/11/12): 535-543.
[5] QU J, COOLEY K M, SHAW A H, et al.Assessment of Wear Coefficients of Nuclear Zirconium Claddings without and with Pre-Oxidation[J]. Wear, 2016, 356: 17-22.
[6] LAZAREVIC S, LU R Y, FAVEDE C, et al.Investigating Grid-to-Rod Fretting Wear of Nuclear Fuel Claddings Using a Unique Autoclave Fretting Rig[J]. Wear, 2018, 412: 30-37.
[7] YIN M G, CAI Z B, YU Y Q, et al.Impact-Sliding Wear Behaviors of 304SS Influenced by Different Impact Kinetic Energy and Sliding Velocity[J]. Tribology International, 2020, 143: 106057.
[8] KIM K T.A Study on the Grid-to-Rod Fretting Wear- Induced Fuel Failure Observed in the 16×16KOFA Fuel[J]. Nuclear Engineering and Design, 2010, 240(4): 756-762.
[9] HU Z P.Developments of Analyses on Grid-to-Rod Fretting Problems in Pressurized Water Reactors[J]. Progress in Nuclear Energy, 2018, 106: 293-299.
[10] KIM K T.The Study on Grid-to-Rod Fretting Wear Models for PWR Fuel[J]. Nuclear Engineering and Design, 2009, 239(12): 2820-2824.
[11] WINTER T, NEU R W, SINGH P M, et al.Coefficient of Friction Evolution with Temperature under Fretting Wear for FeCrAl Fuel Cladding Candidate[J]. Journal of Nuclear Materials, 2019, 520: 140-151.
[12] 黄小波, 宋俊凯, 高玉魁. 核电锆管的表面改性技术[J]. 表面技术, 2016, 45(4): 57-64.
HUANG X B, SONG J K, GAO Y K.Surface Modification Technique of Nuclear Power Zirconium Tube[J]. Surface Technology, 2016, 45(4): 57-64.
[13] 王浩然, 邱长军. 锆合金表面技术的研究进展[J]. 世界有色金属, 2016(18): 46-47.
WANG H R, QIU C J.Research Trend of Zirconium Alloy Surface Technologies[J]. World Nonferrous Metals, 2016(18): 46-47.
[14] GUO X L, LU J Q, LAI P, et al.Understanding the Fretting Corrosion Mechanism of Zirconium Alloy Exposed to High Temperature High Pressure Water[J]. Corrosion Science, 2022, 202: 110300.
[15] LAI P, ZHANG H, ZHANG L F, et al.Effect of Micro-Arc Oxidation on Fretting Wear Behavior of Zirconium Alloy Exposed to High Temperature Water[J]. Wear, 2019, 424: 53-61.
[16] LEE Y H, PARK J H, KIM I H, et al.Enhanced Wear Resistance of CrAl-Coated Cladding for Accident- Tolerant Fuel[J]. Journal of Nuclear Materials, 2019, 523: 223-230.
[17] LI Z Y, CAI Z B, DING Y, et al.Characterization of Graphene Oxide/ZrO2 Composite Coatings Deposited on Zirconium Alloy by Micro-Arc Oxidation[J]. Applied Surface Science, 2020, 506: 144928.
[18] KUMARA C, WANG R, LU R Y, et al.Grid-to-Rod Fretting Wear Study of SiC/SiC Composite Accident- Tolerant Fuel Claddings Using an Autoclave Fretting Bench Test[J]. Wear, 2022, 488: 204172.
[1] 江海霞, 段泽文, 马鹏翔, 等. 核反应堆中锆合金包壳及其表面涂层的微动磨损行为研究进展[J]. 摩擦学学报, 2021, 41(3): 423-436.
JIANG H X, DUAN Z W, MA P X, et al.Research Progress on Fretting Wear Behavior of Fuel Cladding Materials in Nuclear Reactor[J]. Tribology, 2021, 41(3): 423-436.
[19] 王淑祥, 白书欣, 朱利安, 等. 核燃料包壳锆合金表面铬涂层研究进展[J]. 表面技术, 2021, 50(1): 221-231.
WANG S X, BAI S X, ZHU L A, et al.Research Progress of Chromium Coating on Zirconium Alloy for Nuclear Fuel Cladding[J]. Surface Technology, 2021, 50(1): 221-231.
[20] LEE Y H, KIM I H, KIM H K, et al.Role of ZrO2 Oxide Layer on the Fretting Wear Resistance of a Nuclear Fuel Rod[J]. Tribology International, 2020, 145: 106146.
[21] 王俊, 王志国, 蔡振兵, 等. 预氧化锆合金包壳在高温高压水中的微动磨损行为研究[J]. 核动力工程, 2024, 45(5): 142-154.
WANG J, WANG Z G, CAI Z B, et al.Study on Fretting Wear Behavior of Pre-Oxidized Zircaloy Cladding in High Temperature and High Pressure Water[J]. Nuclear Power Engineering, 2024, 45(5): 142-154.
[22] YIN M G, CAI Z B, ZHANG Z X, et al.Effect of Ultrasonic Surface Rolling Process on Impact-Sliding Wear Behavior of the 690 Alloy[J]. Tribology International, 2020, 147: 105600.
[23] 尹美贵, 张磊, 尹海燕. 超声表面滚压工艺影响镍基690合金冲-切磨损行为的研究[J]. 摩擦学学报(中英文), 2024, 44(7): 985-995.
YIN M G, ZHANG L, YIN H Y.Influence of Ultrasonic Surface Rolling on Impact-Shear Wear Behavior of Ni Base 690 Alloy[J]. Tribology, 2024, 44(7): 985-995.
基金
国家自然科学基金青年科学基金(12202465); 四川省自然科学基金青年科学基金(2025ZNSFSC1344); 中央高校基本科研业务费专项资金(25CAFUC04017)