锆合金包壳冲-切耦合磨损行为研究

谢宇杰, 王泞, 李正阳, 宁闯明, 郑裕瀚, 蔡振兵

装备环境工程 ›› 2025, Vol. 22 ›› Issue (10) : 109-122.

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装备环境工程 ›› 2025, Vol. 22 ›› Issue (10) : 109-122. DOI: 10.7643/ issn.1672-9242.2025.10.014
重大工程装备

锆合金包壳冲-切耦合磨损行为研究

  • 谢宇杰1, 王泞1, 李正阳2, 宁闯明1, 郑裕瀚1, 蔡振兵1, *
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Impact-Sliding Wear Behavior of Zirconium Alloy Cladding Tube

  • XIE Yujie1, WANG Ning1, LI Zhengyang2, NING Chuangming1, ZHENG Yuhan1, CAI Zhenbing1, *
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摘要

目的 基于模拟压水堆核电站服役工况,研究切向速度和循环次数对锆合金包壳冲-切耦合磨损行为的影响机理。方法 在冲-切耦合磨损设备上,开展锆合金包壳与支撑格架之间的冲-切耦合磨损试验,研究切向速度与循环次数对包壳磨损界面的动力学响应和损伤行为的影响规律。结果 当切向速度由50 mm/s增至150 mm/s时,磨损率由1.02×104 μm3/J增加至1.256×105 μm3/J,切向运动能量耗散占比从72.85%提升至92.03%,其耗散值显著高于冲击过程。当循环次数增加至2.4×106时,磨损率达到1.106×105 μm3/J。结论 冲-切耦合磨损过程中的能量耗散、磨损体积及其磨损率均与切向速度及循环次数呈正相关。切向运动在材料损伤过程中发挥主导性作用。锆合金燃料包壳的主要磨损机制为剥落、磨粒磨损和氧化磨损。切向速度的增加,加剧了材料的剥落。随着循环次数的增加,包壳表面出现疲劳裂纹的萌生和扩展,磨损机理转变为以疲劳磨损为主。

Abstract

Based on simulating the service conditions of pressurized water reactor (PWR) nuclear power plants, the work aims to investigate the effect mechanisms of tangential velocity and number of cycles on the impact-sliding wear behavior of zirconium alloy cladding. Through the impact-sliding wear testing system, experiments were conducted between zirconium alloy cladding and support grids to investigate the effects of tangential velocity and number of cycles on the dynamic response and damage behavior at the cladding wear interface. As the tangential velocity increased from 50 mm/s to 150 mm/s, the wear rate escalated from 1.02×104 μm3/J to 1.256×105 μm3/J, with tangential motion energy dissipation proportion rising from 72.85% to 92.03%. The energy dissipation through tangential motion significantly exceeded that of impact process. When the number of cycles reached 2.4×106, the wear rate attained 1.106×105 μm3/J. During the impact-sliding wear process, energy dissipation, wear volume, and wear rate all exhibit positive correlations with both tangential velocity and number of cycles. Energy dissipation analysis demonstrates that tangential motion plays a dominant role in the material damage process. The primary wear mechanisms of zirconium alloy cladding include spalling, abrasive wear, and oxidative wear. Increased tangential velocity intensifies material delamination. As the number of cycles accumulates, fatigue cracks initiate and propagate on the cladding surface, shifting the dominant wear mechanism to fatigue-dominated wear.

关键词

锆合金 / 冲-切耦合磨损 / 能量耗散 / 磨损机理 / 燃料包壳 / 磨损率

Key words

zirconium alloy / impact-sliding wear / energy dissipation / wear mechanism / fuel cladding / wear rate

引用本文

导出引用
谢宇杰, 王泞, 李正阳, 宁闯明, 郑裕瀚, 蔡振兵. 锆合金包壳冲-切耦合磨损行为研究[J]. 装备环境工程. 2025, 22(10): 109-122 https://doi.org/10.7643/ issn.1672-9242.2025.10.014
XIE Yujie, WANG Ning, LI Zhengyang, NING Chuangming, ZHENG Yuhan, CAI Zhenbing. Impact-Sliding Wear Behavior of Zirconium Alloy Cladding Tube[J]. Equipment Environmental Engineering. 2025, 22(10): 109-122 https://doi.org/10.7643/ issn.1672-9242.2025.10.014
中图分类号: TL34   

参考文献

[1] CAI Z B, LI Z Y, YIN M G, et al.A Review of Fretting Study on Nuclear Power Equipment[J]. Tribology International, 2020, 144: 106095.
[2] YAN C G, WANG R S, WANG Y L, et al.Effects of Ion Irradiation on Microstructure and Properties of Zirconium Alloys—A Review[J]. Nuclear Engineering and Technology, 2015, 47(3): 323-331.
[3] 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.
[4] SAAD D, BENKHARFIA H, KADOUMA M, et al.Pellet-Cladding Mechanical Interaction Analysis of Heavy Water Fuel Rods under Power Ramps[J]. Annals of Nuclear Energy, 2021, 159: 108320.
[5] LIAO J J, ZHANG J S, ZHANG W, et al.Critical Behavior of Interfacial T-ZrO2 and Other Oxide Features of Zirconium Alloy Reaching Critical Transition Condition[J]. Journal of Nuclear Materials, 2021, 543: 152474.
[6] DUAN Z G, YANG H L, SATOH Y, et al.Current Status of Materials Development of Nuclear Fuel Cladding Tubes for Light Water Reactors[J]. Nuclear Engineering and Design, 2017, 316: 131-150.
[7] KIM T H, KIM S S.Fretting Wear Mechanisms of Zircaloy-4 and Inconel 600 Contact in Air[J]. KSME International Journal, 2001, 15(9): 1274-1280.
[8] ZHANG Y S, LAI J, WANG J Z, et al.Establishment of Fretting Maps of Zr Alloy Cladding Tube Mated with Zr Alloy Dimple in Simulated Primary Water of Pressurized Water Reactor[J]. Tribology International, 2023, 178: 108065.
[9] ZHANG Y S, MING H L, LAI J, et al.Fretting Wear Behaviour of Zr Alloy Cladding Tube under Partial Slip Regime with Different Duration in Simulated Primary Water of PWR[J]. Applied Surface Science, 2022, 605: 154861.
[10] KIM H A, KIM S J.Wear Comfort Properties of ZRC/Al2O3/Graphite-Embedded, Heat-Storage Woven Fabrics for Garments[J]. Textile Research Journal, 2019, 89(8): 1394-1407.
[11] 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.
[12] ZHUANG W H, LAI P, LU H, et al.The Transformation of Fretting Corrosion Mechanism of Zirconium Alloy Tube Mating with 304 Stainless Steel in High Temperature High Pressure Water[J]. Journal of Nuclear Materials, 2023, 577: 154304.
[13] 任全耀, 蒲曾坪, 焦拥军, 等. 高温下锆合金包壳切向微动磨蚀行为研究[J]. 核动力工程, 2022, 43(S2): 82-87.
REN Q Y, PU Z P, JIAO Y J, et al.Study on Tangential Fretting Abrasion Behavior of Zirconium Alloy Cladding at High Temperature[J]. Nuclear Power Engineering, 2022, 43(S2): 82-87.
[14] 焦拥军, 李正阳, 蒲曾坪, 等. 锆合金包壳在微动磨蚀环境下的界面损伤行为[J]. 中国表面工程, 2022, 35(4): 41-49.
JIAO Y J, LI Z Y, PU Z P, et al.Interface Damage Behavior of Zirconium Alloy Cladding under Fretting Corrosion Environment[J]. China Surface Engineering, 2022, 35(4): 41-49.
[15] ATTIA M H, DE PANNEMAECKER A, WILLIAMS G.Effect of Temperature on Tribo-Oxide Formation and the Fretting Wear and Friction Behavior of Zirconium and Nickel-Based Alloys[J]. Wear, 2021, 476: 203722.
[16] WANG J, LI H J, LI Z Y, et al.Effect of Temperature on the Fretting Wear Behavior of Cr-Coated Zircaloy Cladding in High-Temperature Pressurized Water[J]. Journal of Nuclear Materials, 2023, 584: 154516.
[17] PARK Y C, KIM Y H, LEE S J, et al. The Predictions of the Fretting Wear between Supporting Grids and Cladding Tubes of Nuclear Fuel Rod[J]. Key Engineering Materials, 2006, 326/327/328: 1243-1246.
[18] KIM H K, LEE Y H, LEE K H.On the Geometry of the Fuel Rod Supports Concerning a Fretting Wear Failure[J]. Nuclear Engineering and Design, 2008, 238(12): 3321-3330.
[19] 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.
[20] 车龙, 迟百成, 郭策安, 等. 钛合金表面CrN/Cr梯度涂层的微观结构及摩擦磨损性能[J]. 装备环境工程, 2025, 22(4): 1-9.
CHE L, CHI B C, GUO C A, et al.Microstructure and Friction and Wear Properties of CrN/Cr Gradient Coating on Titanium Alloy Surface[J]. Equipment Environmental Engineering, 2025, 22(4): 1-9.
[21] 程杨洋, 钟勇, 贾怡, 等. 激光表面处理对工业级锆基块体非晶合金弯曲变形和缺口韧性的影响[J]. 装备环境工程, 2023, 20(5): 80-89.
CHENG Y Y, ZHONG Y, JIA Y, et al.Effect of Laser Surface Treatment on Bending Deformation and Notch Toughness of Industrial-Grade Zr-Based Bulk Metallic Glasses[J]. Equipment Environmental Engineering, 2023, 20(5): 80-89.
[22] 唐晨, 张伟, 李正阳, 等. 表面纳米化对锆合金微动腐蚀行为的影响[J]. 装备环境工程, 2022, 19(11): 110-118.
TANG C, ZHANG W, LI Z Y, et al.Effect of Surface Nanocrystallization on Fretting Corrosion Behavior of Zirconium Alloy[J]. Equipment Environmental Engineering, 2022, 19(11): 110-118.
[23] CAI Z B, CHEN Z Q, SUN Y, et al.Development of a Novel Cycling Impact-Sliding Wear Rig to Investigate the Complex Friction Motion[J]. Friction, 2019, 7(1): 32-43.
[24] ATTIA H.A Generalized Fretting Wear Theory[J]. Tribology International, 2009, 42(9): 1380-1388.
[25] ZHANG H Y, LU Y H, MA M, et al.Effect of Precipitated Carbides on the Fretting Wear Behavior of Inconel 600 Alloy[J]. Wear, 2014, 315(1/2): 58-67.
[26] CAI Z B, GUAN H D, CHEN Z Q, et al.Impact Fretting Wear Behavior of 304 Stainless Steel Thin-Walled Tubes under Low-Velocity[J]. Tribology International, 2017, 105: 219-228.
[27] YU S J, HU Y, LIU X, et al.Effect of Impact Velocity and Angle on Impact Wear Behavior of Zr-4 Alloy Cladding Tube[J]. Materials, 2022, 15(18): 6371.
[28] PENG J F, DING S Y, HE Z Y, et al.Study on Bi-Directional Composite Fretting Wear Characteristics of Zr-4 Alloy Tube with Different Phase Differences under Temperature Conditions[J]. Tribology International, 2024, 191: 109139.
[29] KIM K T.Applicability of Out-of-Pile Fretting Wear Tests to In-Reactor Fretting Wear-Induced Failure Time Prediction[J]. Journal of Nuclear Materials, 2013, 433(1/2/3): 364-371.
[30] LIN Y W, CAI Z B, CHEN Z Q, et al.Influence of Diameter-Thickness Ratio on Alloy Zr-4 Tube under Low-Energy Impact Fretting Wear[J]. Materials Today Communications, 2016, 8: 79-90.
[31] CAI Z B, LI Z Y, DING Y, et al.Preparation and Impact Resistance Performance of Bionic Sandwich Structure Inspired from Beetle Forewing[J]. Composites Part B: Engineering, 2019, 161: 490-501.
[32] CHEN X D, WANG L W, YANG L Y, et al.Investigation on the Impact Wear Behavior of 2.25Cr-1Mo Steel at Elevated Temperature[J]. Wear, 2021, 476: 203740.
[33] WU S B, CAI Z B, LIN Y, et al.Effect of Abrasive Particle Hardness on Interface Response and Impact Wear Behavior of TC17 Titanium Alloy[J]. Materials Research Express, 2019, 6(1): 016521.
[34] 张晓宇, 任平弟, 张亚非, 等. Incoloy800合金的高温微动磨损特性[J]. 中国有色金属学报, 2010, 20(8): 1545-1551.
ZHANG X Y, REN P D, ZHANG Y F, et al.Fretting Wear Behavior of Incoloy800 Alloy at High Temperature[J]. The Chinese Journal of Nonferrous Metals, 2010, 20(8): 1545-1551.
[35] SILVA C S, HENRIQUES B, NOVAES DE OLIVEIRA A P, et al. Micro-Scale Abrasion and Sliding Wear of Zirconium-Lithium Silicate Glass-Ceramic and Polymer-Infiltrated Ceramic Network Used in Dentistry[J]. Wear, 2020, 448: 203214.
[36] 葛辛辛, 赵南, 杨骏, 等. 温度对编织复合材料层合厚板冲击性能的影响研究[J]. 装备环境工程, 2023, 20(9): 178-184.
GE X X, ZHAO N, YANG J, et al.Effect of Temperature on Impact Resistance of Woven Composite Thick Laminates[J]. Equipment Environmental Engineering, 2023, 20(9): 178-184.
[37] 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.
[38] KIM K T.The Effect of Fuel Rod Supporting Conditions on Fuel Rod Vibration Characteristics and Grid-to-Rod Fretting Wear[J]. Nuclear Engineering and Design, 2010, 240(6): 1386-1391.
[39] ALI BHATTI N, ABDEL WAHAB M.Fretting Fatigue Crack Nucleation: A Review[J]. Tribology International, 2018, 121: 121-138.
[40] TU X M, XIE S H, ZHOU Y Y, et al.Surface Gradient Heterogeneity Induced Tensile Plasticity in a Zr-Based Bulk Metallic Glass through Ultrasonic Impact Treatment[J]. Journal of Non-Crystalline Solids, 2021, 554: 120612.
[41] SINGH K, TIWARI M, MAHATO A.Evolution of Regimes of Wear in Zircaloy-4/Inconel-600 Contact Subjected to Fretting Loading[J]. Tribology International, 2020, 147: 106274.

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