The work aims to improve the service reliability of Zr-4 alloy fuel cladding in high-temperature and high-pressure water environments and provide experimental support for the material selection and structural design of surface protection coatings for accident-tolerant fuel cladding. With Zr-4 alloy as the substrate, NiCr/Mo and NiCrMo/Mo composite coatings were prepared by laser cladding. By adjusting the contents of Cr, Ni, and Mo and combining these compositional modifications with process optimization, coating systems with smooth surfaces, dense microstructures, and good interfacial bonding were successfully obtained. To evaluate the corrosion behavior of the coatings under high-temperature and high-pressure water environments relevant to reactor service conditions, long-term corrosion experiments lasting 100 days were conducted at 360 ℃ and 18.6 MPa in deoxygenated water (DO<10 μg/L) and oxygenated water (DO≈400 µg/L). After corrosion exposure, the surface morphologies, elemental distributions, and mass variations of the coatings were comprehensively characterized by scanning electron microscopy (SEM), energy-dispersive spectroscopy (EDS), and mass-change measurements. The results showed that the Mo interlayer played an important role in improving coating stability. It effectively alleviated the thermal expansion mismatch between the coating and the Zr-4 substrate and suppressed elemental interdiffusion at the interface. As a result, the composite coatings maintained relatively good structural integrity during short-term corrosion. After exposure, the corroded surfaces were mainly covered with nanoscale oxide particles, accompanied by localized selective oxidation, indicating that the coatings retained their overall continuity and did not undergo severe early-stage degradation. During long-term corrosion, all six composite coatings exhibited better corrosion resistance than the Zr-4 substrate, although their performance differed significantly depending on composition and water chemistry. In deoxygenated water, the Ni55Cr40Mo5/Mo coating showed the smallest mass loss, with a 100-day weight-loss rate of -129.24 mg/dm2, indicating the best corrosion resistance among the tested coatings. In contrast, the Ni59Cr40Mo1/Mo coating exhibited the largest mass loss, with a weight-loss rate of -277.64 mg/dm2, and therefore showed the poorest corrosion resistance. This result suggested that an appropriate amount of Mo was beneficial for corrosion protection in a low-oxygen environment. In oxygenated water, the ranking of corrosion performance changed. The Ni60Cr40/Mo coating showed the lowest mass loss, with a 100-day weight-loss rate of -242.00 mg/dm2, while the Ni55Cr40Mo5/Mo coating ranked second, with a value of -257.64 mg/dm2. The Ni59Cr40Mo1/Mo coating again performed the worst, with a weight-loss rate of -306.44 mg/dm2. These results indicated that the optimal coating composition was strongly dependent on the water chemistry, and that coatings performing well in deoxygenated water did not necessarily remain the best in oxygenated conditions. Overall, the work demonstrates that the synergistic design of Mo, Ni, and Cr can significantly enhance the corrosion resistance of laser-cladded coatings in high-temperature water environments. The uniform distribution of oxidation products and the stability of the coating/substrate interface are key factors in maintaining long-term protective performance. These findings indicate that the optimization of protective coatings must consider both compositional design and environmental water chemistry. The results also provide useful experimental evidence for the development of surface protective layers for accident-tolerant fuel cladding, while further verification under longer exposure times and coupled service conditions is still needed.
Key words
Zr-4 alloy /
laser cladding /
NiCr/Mo composite coatings /
high-temperature and high-pressure water corrosion /
long-term corrosion resistance /
interface diffusion
{{custom_sec.title}}
{{custom_sec.title}}
{{custom_sec.content}}
References
[1] TERRANI K A.Accident Tolerant Fuel Cladding Development: Promise, Status, and Challenges[J]. Journal of Nuclear Materials, 2018, 501: 13-30.
[2] YANG J Q, STEINBRÜCK M, TANG C C, et al. Review on Chromium Coated Zirconium Alloy Accident Tolerant Fuel Cladding[J]. Journal of Alloys and Compounds, 2022, 895: 162450.
[3] 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.
[4] ZINKLE S J, TERRANI K A, GEHIN J C, et al.Accident Tolerant Fuels for LWRS: A Perspective[J]. Journal of Nuclear Materials, 2014, 448(1/2/3): 374-379.
[5] 杨红艳, 张瑞谦, 彭小明, 等. 锆合金包壳表面涂层研究进展[J]. 表面技术, 2017, 46(1): 69-77.
YANG H Y, ZHANG R Q, PENG X M, et al.Research Progress Regarding Surface Coating of Zirconium Alloy Cladding[J]. Surface Technology, 2017, 46(1): 69-77.
[6] 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.
[7] 曹国钦, 任莹莹, 仵康康, 等. 轻水堆包壳锆材服役环境下延寿策略及研究进展[J]. 表面技术, 2019, 48(11): 69-81.
CAO G Q, REN Y Y, WU K K, et al.Life Extension Strategy and Research Progress of Zirconium Alloy Cladding Applied to Light Water Reactors[J]. Surface Technology, 2019, 48(11): 69-81.
[8] KOO Y H, YANG J H, PARK J Y, et al.KAERI’s Development of LWR Accident-Tolerant Fuel[J]. Nuclear Technology, 2014, 186(2): 295-304.
[9] CHEN Z P, REN X N, WANG P, et al.Microstructure, Corrosion Resistance and High Temperature Oxidation Properties of AlCrFeNiCu High-Entropy Alloy Coating by High-Speed Laser Cladding[J]. Surface and Coatings Technology, 2024, 491: 131159.
[10] LI Q, SONG P, ZHANG R Q, et al.Corrosion Mechanism and Performance of Cr-Coated Zr-4 Alloy in 360 ℃ Water and 1300 ℃ Steam[J]. Journal of Materials Research and Technology, 2024, 29: 1524-1534.
[11] WANG Y, CHEN B, WANG X Z, et al.Evolution of Cr/Cr2O3 Interface in Cr-Coated Zirconium Alloy in High Temperature Steam[J]. Corrosion Science, 2023, 217: 111099.
[12] QI M, CUI X F, JIN G, et al.Microstructure and Properties of a Multilayered Laser Cladding Al0.2NbTiV0.1W0.5Zr0.3 High-Entropy Alloy Coating on a Zirconium Alloy[J]. Surface and Coatings Technology, 2024, 477: 130299.
[13] WANG Y Y, CHAI L J, ZHANG F L, et al.Laser-Clad Al-Ti-Zr and Cr-Ti-Zr Coatings on Zr Alloy: Composition- Induced Microstructural and Hardness Differences[J]. International Journal of Refractory Metals and Hard Materials, 2022, 109: 105956.
[14] NI N, LOZANO-PEREZ S, JENKINS M L, et al.Porosity in Oxides on Zirconium Fuel Cladding Alloys, and Its Importance in Controlling Oxidation Rates[J]. Scripta Materialia, 2010, 62(8): 564-567.
[15] EL KADIRI H, UTEGULOV Z N, KHAFIZOV M, et al.Transformations and Cracks in Zirconia Films Leading to Breakaway Oxidation of Zircaloy[J]. Acta Materialia, 2013, 61(11): 3923-3935.
[16] ZHANG H L, KIM T, SWARTS J, et al.Nano-Porosity Effects on Corrosion Rate of Zr Alloys Using Nanoscale Microscopy Coupled to Machine Learning[J]. Corrosion Science, 2022, 208: 110660.
[17] KO J, KIM J W, MIN H W, et al.Review of Manufacturing Technologies for Coated Accident Tolerant Fuel Cladding[J]. Journal of Nuclear Materials, 2022, 561: 153562.
[18] HE X D, JIN Y, LIU Z J, et al.Effect of Diffusion Barriers on Steam Oxidation Properties and Interface Evolutions of Cr Coating for Zirconium Alloy at 1200 ℃[J]. Nuclear Engineering and Design, 2024, 429: 113614.
[19] PARK J H, KIM H G, PARK J Y, et al.High Temperature Steam-Oxidation Behavior of Arc Ion Plated Cr Coatings for Accident Tolerant Fuel Claddings[J]. Surface and Coatings Technology, 2015, 280: 256-259.
[20] SIDELEV D V, BLEYKHER G A, KRIVOBOKOV V P, et al.High-Rate Magnetron Sputtering with Hot Target[J]. Surface and Coatings Technology, 2016, 308: 168-173.
[21] BRACHET J C, IDARRAGA-TRUJILLO I, LE FLEM M, et al.Early Studies on Cr-Coated Zircaloy-4 as Enhanced Accident Tolerant Nuclear Fuel Claddings for Light Water Reactors[J]. Journal of Nuclear Materials, 2019, 517: 268-285.
[22] HAN X C, XUE J X, PENG S M, et al.An Interesting Oxidation Phenomenon of Cr Coatings on Zry-4 Substrates in High Temperature Steam Environment[J]. Corrosion Science, 2019, 156: 117-124.
[23] HUANG T, ZHOU Y H, CHEN K, et al.The Oxidation- Dissolution Behavior of Cr-Coated Zr Alloy in High Temperature Water[J]. Journal of Nuclear Materials, 2025, 604: 155504.
[24] MA J J, TAN J, ZHOU Y, et al.Influence of a Mo Interlayer on the Oxidation Behavior of a Cr Coating on a Zr Alloy Substrate[J]. Corrosion Science, 2023, 218: 111192.
[25] YEOM H, MAIER B, JOHNSON G, et al.Development of Cold Spray Process for Oxidation-Resistant FeCrAl and Mo Diffusion Barrier Coatings on Optimized ZIRLO™[J]. Journal of Nuclear Materials, 2018, 507: 306-315.
[26] JIN D L, YANG F, ZOU Z H, et al.A Study of the Zirconium Alloy Protection by Cr3C2-NiCr Coating for Nuclear Reactor Application[J]. Surface and Coatings Technology, 2016, 287: 55-60.
[27] WANG Z, ZHANG Y P, ZHOU S H, et al.Oxidation Resistance of Mo/Cr Bilayer Coating on Zr Alloy in a 1200 ℃ Steam Environment[J]. Corrosion Communications, 2024, 14: 49-57.
[28] LI B, SHEN J J, YANG Z D, et al.Interfacial Evolution in Cr/Mo-Coated Zr Alloys near the Mo-Zr Eutectoid Temperature[J]. Journal of Nuclear Materials, 2025, 603: 155406.
[29] KIM H G, KIM I H, JUNG Y I, et al.Adhesion Property and High-Temperature Oxidation Behavior of Cr-Coated Zircaloy-4 Cladding Tube Prepared by 3D Laser Coating[J]. Journal of Nuclear Materials, 2015, 465: 531-539.
[30] MA H B, ZHANG H L, HU L J, et al.Corrosion Behavior of Cr-Coated Zirconium Alloy Cladding in LiOH/H3BO3- Containing Water at 360 ℃[J]. Corrosion Science, 2023, 222: 111386.
[31] 岳慧芳, 王彦峰, 庞华, 等. Zr-4合金表面Cr涂层长期水腐蚀行为研究[J]. 中国核电, 2024, 17(2): 176-183.
YUE H F, WANG Y F, PANG H, et al.Study on Long- Term Water Corrosion Behaviors of Cr-Coating on Zircaloy-4[J]. China Nuclear Power, 2024, 17(2): 176-183.
[32] BOUVRY B, CHEYMOL G, RAMIANDRISOA L, et al.Multispectral Pyrometry for Surface Temperature Measurement of Oxidized Zircaloy Claddings[J]. Infrared Physics & Technology, 2017, 83: 78-87.
[33] WU X C, WANG D, ZHANG Y P, et al.Isothermal Experiments on Eutectic and Oxidation Reactions of Cr-Coated Zr Alloy Cladding in Steam at 1350 ℃: Behavior, Mechanism and Kinetics[J]. Journal of Alloys and Compounds, 2024, 991: 174562.
Funding
China National Nuclear Corporation Leading Innovation Research Project (CNNC-LCKY-202231)
PDF(28786 KB)
