极端工况下金属型核燃料棒制备技术：材料创新与工艺优化研究进展

王延; 周弋凇; 谢旭峰; 胡红军; 钟韬

doi:10.3969/j.issn.1674-6457.2026.06.021

PDF(35709 KB)

精密成形工程 ›› 2026, Vol. 18 ›› Issue (6) : 228-259. DOI: 10.3969/j.issn.1674-6457.2026.06.021

难熔金属成形

极端工况下金属型核燃料棒制备技术：材料创新与工艺优化研究进展

王延¹, 周弋凇¹, 谢旭峰¹, 胡红军^1,*, 钟韬^2,*

作者信息 +

Fabrication Technology of Metallic Nuclear Fuel Rods under Extreme Conditions: Research Progress in Material Innovation and Process Optimization

WANG Yan¹, ZHOU Yisong¹, XIE Xufeng¹, HU Hongjun^1,*, ZHONG Tao^2,*

Author information +

文章历史 +

摘要

随着全球能源结构向低碳化转型,核能因其高能量密度与低碳排放特性成为关键能源形式。核燃料作为核反应堆的核心组件,其性能与安全性直接影响反应堆效率及可靠性。本文系统综述了金属型核燃料棒制备加工技术的研究现状与挑战,分别介绍了核燃料芯体与包壳材料的开发与应用、芯体-包壳相互作用以及先进制备工艺。研究表明,金属型燃料通常以铀为核心,其中合金元素对微观组织演变和相稳定性具有显著影响。芯体与包壳的紧密配合可能会影响两者相互作用。此外,在制备工艺中,铸造通常用于前期;共挤压技术可实现多层结构的高效成型,但仍需优化工艺参数、提高界面结合性能;轧制技术应用广泛,可以获得高质量的金属型燃料零件。在包壳材料方面,锆合金因低中子吸收截面与优异的耐腐蚀性仍为主要应用材料,但其高温氢脆与氧化问题亟待解决;SiC等新型包壳材料展现出事故容错潜力,但其抗辐照性能仍需进一步提升。在芯体-包壳相互作用机制中,化学扩散与热应力失配是导致界面失效的主因,需通过扩散屏障设计与材料匹配进行工艺优化。目前针对表面改性的研究持续深入,增材制造技术也为复杂结构燃料制备提供了新途径。

Abstract

As the global energy structure transitions toward low-carbonization, nuclear energy has emerged as a critical energy source owing to its high energy density and low-carbon emission characteristics. As the core component of nuclear reactors, nuclear fuel directly affects reactor efficiency and reliability in terms of its performance and safety. This paper systematically reviews the research status and challenges of fabrication and processing technologies for metallic nuclear fuel rods, and introduces the development and application of nuclear fuel pellets and cladding materials, pellet-cladding interaction, as well as advanced fabrication processes. Studies indicate that metallic fuels are generally uranium-based, in which alloying elements exert significant effects on microstructural evolution and phase stability. Tight coupling between pellets and cladding may influence their mutual interaction. In terms of fabrication processes, casting is commonly adopted in preliminary stages; co-extrusion enables efficient forming of multilayer structures yet requires further optimization of process parameters and improvement of interfacial bonding performance; rolling technology is widely applied to fabricate high-quality metallic fuel components. For cladding materials, zirconium alloys remain the dominant choice due to their low neutron absorption cross section and excellent corrosion resistance, whereas issues such as high-temperature hydrogen embrittlement and oxidation need urgent solutions. Innovative cladding materials including silicon carbide (SiC) exhibit accident-tolerant potential, yet their radiation resistance requires further enhancement. Within the pellet-cladding interaction mechanism, chemical diffusion and thermal stress mismatch are the primary causes of interfacial failure, which requires process optimization via diffusion barrier design and material matching. Current research on surface modification is continuously advancing, and additive manufacturing technologies provide new approaches for fabricating fuels with complex structures.

导出引用

王延, 周弋凇, 谢旭峰, 胡红军, 钟韬. 极端工况下金属型核燃料棒制备技术：材料创新与工艺优化研究进展[J]. 精密成形工程. 2026, 18(6): 228-259 https://doi.org/10.3969/j.issn.1674-6457.2026.06.021

WANG Yan, ZHOU Yisong, XIE Xufeng, HU Hongjun, ZHONG Tao. Fabrication Technology of Metallic Nuclear Fuel Rods under Extreme Conditions: Research Progress in Material Innovation and Process Optimization[J]. Journal of Netshape Forming Engineering. 2026, 18(6): 228-259 https://doi.org/10.3969/j.issn.1674-6457.2026.06.021

中图分类号： TL341

参考文献

[1] 林继铭, 田瑞峰, 李凤臣, 等. 液态重金属反应堆工程关键技术研究综述[J]. 核技术, 2025, 48(7): 53-68.
LIN J M, TIAN R F, LI F C, et al.Review of Key Engineering Technologies for Liquid Heavy Metal Cooled Reactors[J]. Nuclear Techniques, 2025, 48(7): 53-68.
[2] 张羽廷. 铀锆合金燃料制备技术研究[D]. 合肥: 中国科学技术大学, 2017.
ZHANG Y T.Manufacturing Methods of U-Zr Metallic Fuel[D]. Hefei: University of Science and Technology of China, 2017.
[3] 冯伟, 刘一哲, 任媛媛, 等. 一体化快堆金属燃料发展现状及研发规划[J]. 原子能科学技术, 2025, 59(S1): 34-40.
FENG W, LIU Y Z, REN Y Y, et al.Development Status and Plan of Metal Fuel of Integrated Fast Reactor[J]. Atomic Energy Science and Technology, 2025, 59(S1): 34-40.
[4] 李静, 李峻宏, 冯伟, 等. 一体化快堆包壳用ODS合金的研制[J]. 原子能科学技术, 2025, 59(S1): 148-158.
LI J, LI J H, FENG W, et al.Preliminary Study on ODS Alloy for Integrated Fast Reactor[J]. Atomic Energy Science and Technology, 2025, 59(S1): 148-158.
[5] 尹传凯, 王春连, 游超, 等. 中国锆矿资源特征、矿床类型、关键应用及找矿远景[J]. 中国地质, 2024, 51(6): 1930-1945.
YIN C K, WANG C L, YOU C, et al.Zircon Resource Characteristics, Deposit Types, Key Applications and Prospecting Prospects in China[J]. Geology in China, 2024, 51(6): 1930-1945.
[6] 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.
[7] JHA S K, DIXIT S, CHETAN K, et al.Co-Extrusion of Zircaloy-2 and Zr-Sn Alloy for Double Clad Tube Manufacturing: Numerical Simulation and Experimental Validation[J]. Journal of Manufacturing Processes, 2019, 39: 18-25.
[8] LIU Y M, WANG F, LIU H T, et al.Compatibilities and Interfacial Diffusions between U-10Zr Alloy and Zircaloy-4 Alloy[J]. Journal of Nuclear Materials, 2016, 473: 223-228.
[9] 刘文博. FeCrAl(-Zr)合金变形和热处理过程中的组织演变及力学性能研究[D]. 长沙: 中南大学, 2023.
LIU W B.Microstructure Evolution and Mechanical Properties of FeCrAl(-Zr) Alloy during the Deformation and Annealing[D]. Changsha: Central South University, 2023.
[10] AZEVEDO C R F. Selection of Fuel Cladding Material for Nuclear Fission Reactors[J]. Engineering Failure Analysis, 2011, 18(8): 1943-1962.
[11] 潘荣剑, 汤爱涛, 吴璐, 等. 锆合金包壳材料堆内行为的研究进展[J]. 材料科学, 2019(9): 861-871.
PAN R J, TANG A T, WU L, et al.Recent Progress on In-Pile Behavior of Zirconium Alloy Cladding Materials[J]. Material Sciences, 2019(9): 861-871.
[12] ZHU W H, WANG R S, SHU G G, et al.First-Principles Study of Different Polymorphs of Crystalline Zirconium Hydride[J]. The Journal of Physical Chemistry C, 2010, 114(50): 22361-22368.
[13] WANG F, GONG H R.Mechanical and Structural Stability of Zirconium Dihydride[J]. International Journal of Hydrogen Energy, 2012, 37(12): 9688-9695.
[14] LUMLEY S C, GRIMES R W, MURPHY S T, et al.The Thermodynamics of Hydride Precipitation: The Importance of Entropy, Enthalpy and Disorder[J]. Acta Materialia, 2014, 79: 351-362.
[15] STOJILOVIC N, BENDER E T, RAMSIER R D.Oxidation of Zircaloy-4 by Oxygen and the Production of Water[J]. Journal of Nuclear Materials, 2006, 348(1/2): 79-86.
[16] YILMAZBAYHAN A, BREVAL E, MOTTA A T, et al.Transmission Electron Microscopy Examination of Oxide Layers Formed on Zr Alloys[J]. Journal of Nuclear Materials, 2006, 349(3): 265-281.
[17] QIN W, NAM C, LI H L, et al.Tetragonal Phase Stability in ZrO₂ Film Formed on Zirconium Alloys and Its Effects on Corrosion Resistance[J]. Acta Materialia, 2007, 55(5): 1695-1701.
[18] 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.
[19] MALLIPUDI V, VALANCE S, BERTSCH J.Meso- Scale Analysis of the Creep Behavior of Hydrogenated Zircaloy-4[J]. Mechanics of Materials, 2012, 51: 15-28.
[20] MOON J H, CANTONWINE P E, ANDERSON K R, et al.Characterization and Modeling of Creep Mechanisms in Zircaloy-4[J]. Journal of Nuclear Materials, 2006, 353(3): 177-189.
[21] SARKAR A, BOOPATHY K, EAPEN J, et al.Creep Behavior of Hydrogenated Zirconium Alloys[J]. Journal of Materials Engineering and Performance, 2014, 23(10): 3649-3656.
[22] 刘鹏, 李松发, 戚雄飞, 等. 研究堆用铝合金包壳堆内平均腐蚀速率评估方法研究[J]. 核科学与工程, 2022, 42(3): 504-509.
LIU P, LI S F, QI X F, et al.Study on the Evaluation Method for the Average Corrosion Rate of Aluminum Alloy Cladding for Research Reactor[J]. Nuclear Science and Engineering, 2022, 42(3): 504-509.
[23] 郁金南. 材料辐照效应[M]. 北京: 化学工业出版社, 2007: 2-3.
YU J N.Material Irradiation Effect[M]. Beijing: Chemical Industry Press, 2007: 2-3.
[24] 胡凌, 郑航, 冯琦杰, 等. 长期中子辐照 Al-Mg-Si合金的压缩力学行为[J]. 爆炸与冲击, 2019, 39(12): 44-53.
HU L, ZHENG H, FENG Q J, et al.Mechanical Behavior of Long-Term Neutron-Irradiated Al-Mg-Si Alloy under Compression[J]. Explosion and Shock Waves, 2019, 39(12): 44-53.
[25] LOBB R C, JONES R B.The Influence of Iodine Vapour on Creep Rupture Properties of 20%Cr/25%Ni/Nb Stabilised Stainless Steel[J]. Journal of Nuclear Materials, 1976, 59(3): 280-292.
[26] LOBB R C.The Effect of Iodine Vapour on Creep Rupture Properties of Nitrided 20% Cr/25% Ni/Nb/1.5 Ti Stainless Steel[J]. Journal of Nuclear Materials, 1978, 74(2): 212-220.
[27] ANTILL J E, PEAKALL K A, SMART E F.Corrosion of Stainless Steel in the Presence of Caesium[J]. Journal of Nuclear Materials, 1975, 56(1): 47-60.
[28] CHO H S, KIMURA A, UKAI S, et al.Corrosion Properties of Oxide Dispersion Strengthened Steels in Super-Critical Water Environment[J]. Journal of Nuclear Materials, 2004, 329: 387-391.
[29] MARÉCHAL L, LESAGE B, HUNTZ A M, et al. Oxidation Behavior of ODS Fe-Cr-Al Alloys: Aluminum Depletion and Lifetime[J]. Oxidation of Metals, 2003, 60(1): 1-28.
[30] LIPKINA K, HALLATT D, GEIGER E, et al.A Study of the Oxidation Behaviour of FeCrAl-ODS in Air and Steam Environments up to 1 400 ℃[J]. Journal of Nuclear Materials, 2020, 541: 152305.
[31] TAKAYA S, FURUKAWA T, MÜLLER G, et al. Al-Containing ODS Steels with Improved Corrosion Resistance to Liquid Lead-Bismuth[J]. Journal of Nuclear Materials, 2012, 428(1/2/3): 125-130.
[32] KIMURA A, KASADA R, IWATA N, et al.Development of Al Added High-Cr ODS Steels for Fuel Cladding of Next Generation Nuclear Systems[J]. Journal of Nuclear Materials, 2011, 417(1/2/3): 176-179.
[33] 翟剑晗. FeCrAl先进不锈钢包壳管材力学性能研究[J]. 冶金与材料, 2024, 16(7): 181-183.
ZHAI J H.Study on Mechanical Properties of FeCrAl Advanced Stainless Steel Cladding Pipe[J]. Metallurgical and Materials, 2024, 16(7): 181-183.
[34] STASIAK T, JASIŃSKI J J, ZIELIŃSKI M, et al. Mechanical Properties, 475 ℃ Embrittlement and Irradiation Resistance of FeCrAl-Y₂O₃ ODS Alloy with Ti and V Alloying Additions[J]. Journal of Nuclear Materials, 2025, 617: 156131.
[35] 石浩江, 李权, 孙永铎, 等. SiC_f/SiC复合材料在核能领域应用设计的历史回顾、挑战及发展[J]. 复合材料学报, 2025, 42(11): 6130-6151.
SHI H J, LI Q, SUN Y D, et al.Historical Review, Challenges and Development of SiC_f/SiC Composite for Nuclear Applications[J]. Acta Materiae Compositae Sinica, 2025, 42(11): 6130-6151.
[36] LEE Y, KAZIMI M S.A Structural Model for Multi-Layered Ceramic Cylinders and Its Application to Silicon Carbide Cladding of Light Water Reactor Fuel[J]. Journal of Nuclear Materials, 2015, 458: 87-105.
[37] LI B.Pellet Cladding Mechanical Interactions of Ceramic Claddings Fuels under Light Water Reactor Conditions[D]. Columbia, South Carolina, USA: University of South Carolina, 2013.
[38] STEMPIEN J D, CARPENTER D M, KOHSE G, et al.Characteristics of Composite Silicon Carbide Fuel Cladding after Irradiation under Simulated PWR Conditions[J]. Nuclear Technology, 2013, 183(1): 13-29.
[39] 卓卫乾. 燃料包壳化学反应和相优化的U-Zr基金属燃料掺杂及性能改善研究[J]. 核动力工程, 2023, 44(S1): 158-162.
ZHUO W Q.Research on U-Zr-based Metallic Fuel Additives and Performance Improvement for Fuel-Cladding Chemical Interaction and Phase Optimization[J]. Nuclear Power Engineering, 2023, 44(S1): 158-162.
[40] 庞晓轩, 尹昌耕, 沈保罗, 等. U-Mo合金与Nb的互扩散行为[J]. 原子能科学技术, 2008, 42(7): 613-617.
PANG X X, YIN C G, SHEN B L, et al.Diffusion Behavior of U-Mo Alloy with Nb[J]. Atomic Energy Science and Technology, 2008, 42(7): 613-617.
[41] QI Y M, ZHANG C P, HE T W, et al.Irradiation Dose-Dependent Mechanical Response of Titanium through Point Defect Accumulation[J]. Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, 2025, 569: 165899.
[42] CAO Z Q, DING Y F, ZHONG Q H, et al.Defect Evolution in Fe Using In-Situ Fe-H-He Triple-Beam Simultaneous Irradiation under Two Orders of Magnitude of Dose Rates with Constant H/Dpa and He/Dpa Ratios[J]. Journal of Nuclear Materials, 2025, 617: 156157.
[43] ZHU H P, ZHANG Z M, WANG J Q, et al.The Role of Microstructural Evolution in Irradiation Hardening of Alloy 718 under Low Dose Proton Irradiation[J]. Journal of Nuclear Materials, 2025, 606: 155644.
[44] 姚钦文. 铀钼合金燃料辐照肿胀行为综述[J]. 科技创新与应用, 2023, 13(25): 6-10.
YAO Q W.Overview of Radiation Swelling Behavior of Uranium-Molybdenum Alloy Fuel[J]. Technology Innovation and Application, 2023, 13(25): 6-10.
[45] MA J J, HAN X F, CAI X X, et al.High-Temperature Mechanical and Dynamical Properties of γ-(U, Zr) Alloys[J]. Materials, 2023, 16(7): 2623.
[46] XIAO H X, LONG C S, TIAN X F, et al.Atomistic Simulations of the Small Xenon Bubble Behavior in U-Mo Alloy[J]. Materials & Design, 2015, 74: 55-60.
[47] 曹蓓, 白文秀, 姚欢, 等. 核燃料包壳用FeCrAl合金高温-辐照蠕变行为的分子动力学模拟[J]. 南方能源建设, 2025, 12(4): 132-141.
CAO B, BAI W X, YAO H, et al.Molecular Dynamics Simulation of Thermal and Irradiation Creep Behavior of FeCrAl Alloy for Nuclear Fuel Cladding[J]. Southern Energy Construction, 2025, 12(4): 132-141.
[48] HIREL P.Atomsk: A Tool for Manipulating and Converting Atomic Data Files[J]. Computer Physics Communications, 2015, 197: 212-219.
[49] 秦凯文, 杨波, 王子鸣, 等. 不同类型核燃料对热管冷却反应堆燃耗性能的影响[J]. 强激光与粒子束, 2022, 34(12): 126001.
QIN K W, YANG B, WANG Z M, et al.Influence of Different Types of Nuclear Fuel on Burnup Performance of Heat Pipe Cooled Reactor[J]. High Power Laser and Particle Beams, 2022, 34(12): 126001.
[50] 费敬然, 司胜义, 陈其昌. 金属燃料变形多物理效应分析[J]. 核动力工程, 2017, 38(1): 67-71.
FEI J R, SI S Y, CHEN Q C.Multi-Physics Analysis of Metallic Fuel Deformation by Numerical Fuel Rod Code YUAN[J]. Nuclear Power Engineering, 2017, 38(1): 67-71.
[51] SALVATO D, PAAREN K M, HIRSCHHORN J A, et al.The Effect of Temperature and Burnup on U-10Zr Metallic Fuel Chemical Interaction with HT-9: A SEM-EDS Study[J]. Journal of Nuclear Materials, 2024, 591: 154928.
[52] 马涛, 郭成初, 陈铭, 等. 预凝胶法碳热还原制备UO₂核燃料微球[J]. 材料科学与工程学报, 2018, 36(6): 863-867.
MA T, GUO C C, CHEN M, et al.Preliminary Fabrication of UO₂ Microspheres by Carbothermic Reduction and Pregelation Process[J]. Journal of Materials Science and Engineering, 2018, 36(6): 863-867.
[53] 郭鸿豪. U-X(Mo, Zr)和U-Mo-Zr合金在辐照条件下的调幅分解和气泡生长的相场法研究[D]. 厦门: 厦门大学, 2022.
GUO H H.Phase-Field Study on Spinodal Decomposition and Gas Bubble Growth of U-X(Mo, Zr) and U-Mo-Zr Alloy under Irradiation[D]. Xiamen: Xiamen University, 2022.
[54] 李林阳. 辐照条件下部分金属型铀基核燃料与锆基包壳材料的相图计算[D]. 厦门: 厦门大学, 2017.
LI L Y.Phase Diagram Calculation of Partial Metal Uranium-Based Nuclear Fuel and Zirconium-Based Cladding Material under Irradiation[D]. Xiamen: Xiamen University, 2017.
[55] RICHMOND W J, ISSEROW S.Casting of Zircaloy- Uranium Cores for HWCTR Driver Tubes[J]. Nuclear Applications, 1967, 3(5): 294-297.
[56] LÓPEZ M, PICCHETTI B, TABOADA H. Influence of Temperature and Compressive Stress on the UMo/Zry-4 Interdiffusion Layer[J]. Progress in Nuclear Energy, 2017, 94: 101-105.
[57] WANG Y C, MILLER B D, HARP J M, et al.Transmission Electron Microscopy Characterization of the Fuel- Cladding Chemical Interactions in HT9 Cladded U-10Zr Fuel[J]. Journal of Nuclear Materials, 2022, 572: 153990.
[58] MATTEY R, OZALTUN H.Unraveling the Mechanics: Computational Modeling of Residual Stresses in U10Mo Fuel with Aluminum Cladding[R]. Idaho National Laboratory (INL), Idaho Falls Idaho, 2023.
[59] ZHANG Y T, WANG X, LIU P C, et al.Microstructural Characterization and Growth Kinetics of the Reaction Layer in U-10wt% Zr/Zircaloy-4 Diffusion Couples[J]. Rare Metal Materials and Engineering, 2018, 47(9): 2675-2681.
[60] CHEN X D, XIE Z X, MAO X X, et al.Modeling of Zirconium Atom Redistribution and Phase Transformation Coupling Behaviors in U-10Zr-Based Helical Cruciform Fuel Rods under Irradiation[J]. Metals, 2024, 14(7): 745.
[61] KAUTZ E J, SONG M, RIECHERS S, et al.Influence of Processing on Secondary Phase Formation and Microstructural Evolution at U-10Mo Alloy and Zr Interlayer Interfaces[J]. Journal of Alloys and Compounds, 2023, 969: 172074.
[62] 王启明, 严晓青, 霍永忠, 等. 板形弥散型核燃料元件芯体与包壳界面力学行为研究[C]// 中国力学学会, 郑州大学.中国力学学会学术大会2009论文摘要集. 复旦大学力学与工程科学系, 2009: 176.
WANG Q M, YAN X Q, HUO Y Z, et al.Study on the Mechanical Behavior of the Interface between the Core and the Cladding of a Plate-type Dispersed Nuclear Fuel Element[C]// Academic Congress of Chinese Society of Mechanics, Zhengzhou University. Abstracts Collection of the 2009 Academic Conference of Chinese Society of Theoretical and Applied Mechanics. Department of Mechanics and Engineering Science, Fudan University, 2009: 176.
[63] 董颖璇, 吕俊男, 李群. 颗粒团聚行为对弥散型核燃料芯体失效的影响分析[J]. 原子能科学技术, 2024, 58(4): 868-877.
DONG Y X, LYU J N, LI Q.Analysis of Particle Agglomeration Effect on Failure of Dispersion Nuclear Fuel Meat[J]. Atomic Energy Science and Technology, 2024, 58(4): 868-877.
[64] 翁琳, 吴爱中, 张镇国, 等. 颗粒团聚分布的复合材料数值模型[J]. 精密成形工程, 2024, 16(4): 79-86.
WENG L, WU A Z, ZHANG Z G, et al.Numerical Model of Composite Materials with Agglomerated Particles[J]. Journal of Netshape Forming Engineering, 2024, 16(4): 79-86.
[65] JUE J F, KEISER D D, MILLER B D, et al.Effects of Irradiation on the Interface between U-Mo and Zirconium Diffusion Barrier[J]. Journal of Nuclear Materials, 2018, 499: 567-581.
[66] PARK J Y, KIM K, SONG H, et al.Remote Injection Casting Process with Reduced Material Loss for Fabrication of Metallic Fuels[J]. Progress in Nuclear Energy, 2021, 132: 103595.
[67] IRUKUVARGHULA S, AHN S, MCDEAVITT S M.Decomposition of the γ Phase in As-Cast and Quenched U-Zr Alloys[J]. Journal of Nuclear Materials, 2016, 473: 206-217.
[68] KIM J H, SONG H, KIM H T, et al.Development of a New Casting Method to Fabricate U-Zr Alloy Containing Minor Actinides[J]. Journal of Radioanalytical and Nuclear Chemistry, 2014, 299(1): 103-109.
[69] HA S J, KIM K H, PARK J Y, et al.Characterization of Reused U-10Zr Heel Residue Containing Rare-Earth Elements through Surface Treatment[J]. Journal of Radioanalytical and Nuclear Chemistry, 2018, 316(3): 1157-1163.
[70] LAVENDER C A, PAXTON D M, SMITH M T, et al.Concept Feasibility Report for Using Co-extrusion to Bond Metals to Complex Shapes of U-10Mo[R]. Pacific Northwest National Laboratory (PNNL), Richland Washington, 2013.
[71] CRAWFORD D C, PORTER D L. Applying U.S. Metal Fuel Experience to New Fuel Designs for Fast Reactors[J]. Progress in Nuclear Energy, 2024, 171: 105135.
[72] LEE U J, KO Y M, KIM T K, et al.The Effect of Core Designs and Extrusion Speeds on the Coextruded Zr- U/Zr-Nb Fuel Rod[C]// Transactions of the Korean Nuclear Society Spring Meeting Chuncheon, Korea, 2006: 25-26.
[73] FIELDING R, MACKOWIAK B, DEXTER G, et al.Co-Extrusion of Zr Lined U-Zr Alloy Characterization Report[R]. Idaho National Laboratory (INL), Idaho Falls Idaho, 2006.
[74] ARMIJO J S, ROSENBAUM H S, WILLIAMS C D. Method for Making Fuel Cladding Having Zirconium Barrier Layers and Inner Liners: US5383228[P].1995- 01-17.
[75] PASQUALINI E E, ROBINSON A B, PORTER D L, et al.Fabrication and Testing of U-7Mo Monolithic Plate Fuel with Zircaloy Cladding[J]. Journal of Nuclear Materials, 2016, 479: 402-410.
[76] LI L, FORTIER A, RAMIREZ-TAMAYO D, et al.Minimizing Thickness Variation in Monolithic U-10Mo Fuel Foil and Zr Interlayer during Hot Rolling: A Microstructure-Based Finite Element Method Analysis[J]. Materials Today Communications, 2022, 32: 103910.
[77] BEAN C H, MACHEREY R E, LINDGREN J R.Roll Cladding Uranium-Zirconium and Uranium-Zirconium- Niobium Alloys with Zircaloy-2 for Plate-type Fuel Elements[R]. University of North Texas Libraries, UNT Digital Library, Lemont Illinois, 1958.
[78] BEAN C H, MACHEREY R E.Roll Cladding of Uranium-Niobium Alloys for Plate Type Fuel Elements[R]. Argonne National Laboratory (ANL), Lemont Illinois, 1953.
[79] 张丽英, 秦国鹏, 尹富斌. 核燃料零部件的金属3D打印制造技术初步研究[C]// 四川省机械工程学会. 四川省机械工程学会第三届学术年会论文集. 中核建中核燃料元件有限公司, 2018: 188-193.
ZHANG L Y, QIN G P, YIN F B.Primary Research on Metal 3D Printing Manufacturing Technology of Fuel Assembly Parts[C]// Sichuan Mechanical Engineering Society. Proceedings of the Third Annual Conference of Sichuan Mechanical Engineering Society. CNNC Jianzhong Nuclear Fuel Element Co., Ltd., 2018: 188-193.
[80] NELSON A T.Features that Further Performance Limits of Nuclear Fuel Fabrication: Opportunities for Additive Manufacturing of Nuclear Fuels[R]. Oak Ridge National Laboratory (ORNL), Oak Ridge Tennessee, 2019.
[81] 邓话, 秦国鹏. 核燃料零部件的金属增材制造技术研发[J]. 中国核电, 2020, 13(6): 769-773.
DENG H, QIN G P.The Research on Metal Additive Manufacturing Technology of Nuclear Fuel Parts[J]. China Nuclear Power, 2020, 13(6): 769-773.
[82] 廉超, 郭林, 常振旗. 3D打印制备核素分布高度均匀的弥散型陶瓷核燃料芯块[J]. 中国科学技术大学学报, 2020, 50(4): 436-441.
LIAN C, GUO L, CHANG Z Q.Fabrication of Dispersion Ceramic Nuclear Fuel Pellets with a Highly Uniform Distribution of Nuclide by 3D Printing[J]. Journal of University of Science and Technology of China, 2020, 50(4): 436-441.
[83] LIAO H Y, ZHANG T, LI C X, et al.Advancements in the Research of High-Temperature Gas-Cooled Reactor Fuel via Additive Manufacturing Techniques[J]. International Journal of Advanced Nuclear Reactor Design and Technology, 2024, 6(1): 14-20.
[84] WANG Y, WANG L H, SHANG L L, et al.Fiber Texture-Dependent Oxidation Behaviour of Cr-Coated Zirconium Alloy in High Temperature Steam[J]. Corrosion Science, 2022, 205: 110449.
[85] 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.
[86] 徐向空. 锆合金表面抗高温氧化Cr-Si涂层的制备及性能研究[D]. 重庆: 重庆理工大学, 2024.
XU X K.Preparation and Performance Study of High-Temperature Oxidation Resistance Cr-Si Coating on Zirconium Alloy Surface[D]. Chongqing: Chongqing University of Technology, 2024.
[87] CHEN H, WANG X M, ZHANG R Q.Application and Development Progress of Cr-Based Surface Coatings in Nuclear Fuel Element: I. Selection, Preparation, and Characteristics of Coating Materials[J]. Coatings, 2020, 10(9): 808.
[88] TANG C C, STUEBER M, SEIFERT H J, et al.Protective Coatings on Zirconium-Based Alloys as Accident-Tolerant Fuel (ATF) Claddings[J]. Corrosion Reviews, 2017, 35(3): 141-165.
[89] 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.
[90] 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.
[91] TERRANI K A.Accident Tolerant Fuel Cladding Development: Promise, Status, and Challenges[J]. Journal of Nuclear Materials, 2018, 501: 13-30.
[92] BAI G H, CHEN Z L, ZHANG Y W, et al.Research Progress of Coating on Zirconium Alloy for Nuclear Fuel Cladding[J]. Rare Metal Materials and Engineering, 2017, 46: 2035-20.
[93] 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: 69-77.
[94] ZHANG W, TANG R, YANG Z B, et al.Preparation, Structure, and Properties of High-Entropy Alloy Multilayer Coatings for Nuclear Fuel Cladding: A Case Study of AlCrMoNbZr/(AlCrMoNbZr)N[J]. Journal of Nuclear Materials, 2018, 512: 15-24.
[95] OH J M, KIM D, YOO S, et al.Comparative Study of Cr-Coating and Zr-Lining Method on Inner Surface of HT9 Cladding to Mitigate Fuel-Cladding Chemical Interaction[J]. Nuclear Engineering and Design, 2024, 421: 113102.
[96] 朱林丹, 肖华强, 任丽蓉, 等. 锆合金表面高温抗氧化涂层的研究进展[J]. 中国有色金属学报, 2024, 34(9): 2869-2894.
ZHU L D, XIAO H Q, REN L R, et al.Research Progress of High-Temperature Oxidation-Resistant Coatings on Surface Zirconium Alloy[J]. The Chinese Journal of Nonferrous Metals, 2024, 34(9): 2869-2894.
[97] 朱咸勇, 丁振宇, 马国政, 等. 三元MAX相层状陶瓷材料高温摩擦学性能研究进展[J]. 材料导报, 2022, 36(7): 69-79.
ZHU X Y, DING Z Y, MA G Z, et al.Research Progress on High Temperature Tribological Properties of Ternary MAX Phase Layered Ceramic Materials[J]. Materials Review, 2022, 36(7): 69-79.
[98] AZINA C, MRÁZ S, GRECZYNSKI G, et al. Oxidation Behaviour of V₂AlC MAX Phase Coatings[J]. Journal of the European Ceramic Society, 2020, 40(13): 4436-4444.
[99] LI W T, WANG Z Y, SHUAI J T, et al.A High Oxidation Resistance Ti₂AlC Coating on Zirlo Substrates for Loss-of-Coolant Accident Conditions[J]. Ceramics International, 2019, 45(11): 13912-13922.
[100] KIM H G, KIM I H, JUNG Y I, et al.Out-of-Pile Performance of Surface-Modified Zr Cladding for Accident Tolerant Fuel in LWRS[J]. Journal of Nuclear Materials, 2018, 510: 93-99.
[101] HE L X, LIU C H, LIN J H, et al.Microstructure, Oxidation and Corrosion Properties of FeCrAl Coatings with Low Al Content Prepared by Magnetron Sputtering for Accident Tolerant Fuel Cladding[J]. Journal of Nuclear Materials, 2021, 551: 152966.
[102] ALAT E, MOTTA A T, COMSTOCK R J, et al.Multilayer (TiN, TiAlN) Ceramic Coatings for Nuclear Fuel Cladding[J]. Journal of Nuclear Materials, 2016, 478: 236-244.
[103] ALAT E, MOTTA A T, COMSTOCK R J, et al.Ceramic Coating for Corrosion (C3) Resistance of Nuclear Fuel Cladding[J]. Surface and Coatings Technology, 2015, 281: 133-143.
[104] 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.
[105] 李刚, 王辉, 吴松岭. 先进核燃料用ZrC涂层材料的研究进展[J]. 核科学与技术, 2020(1): 42-51.
LI G, WANG H, WU S L.Research Progress of ZrC Coating Materials for Advanced Nuclear Fuels[J]. Nuclear Science and Technology, 2020(1): 42-51.
[106] LIN J C, CHEN Y, MA D C, et al.Achieving Nearly-Homogeneous Joint of ZRC-SiC Composite via Liquid-Film Assisted Transformation Strategy Enabled by Pulsed Current[J]. Journal of Materials Science & Technology, 2026, 243: 192-205.
[107] WANG Y J, LI X Q, ZHU W J, et al.Low-Temperature and Pressureless Fabrication of High-Purity SiC-ZrC Ceramics Free of Residual Melts via Reactive Melt Infiltration[J]. Journal of Alloys and Compounds, 2025, 1045: 184689.
[108] SUN Z Q, BEI H B, YAMAMOTO Y.Microstructural Control of FeCrAl Alloys Using Mo and Nb Additions[J]. Materials Characterization, 2017, 132: 126-131.
[109] ZHOU R N, WANG F M, XU K, et al.Effect of Molybdenum Addition on Oxidation Behavior and Secondary Protection Mechanism of FeCrAl Coatings[J]. Materials Characterization, 2023, 204: 113221.
[110] ZHU P Z, WANG J J, HUANG W J, et al.Insight into the Structure and Protective Performance of Mo/FeCrAl Bilayer Coatings on Zry-4 Substrates in Hydrothermal Corrosion and High-Temperature Steam Environment[J]. Surface and Coatings Technology, 2025, 497: 131733.
[111] ZHOU X, WANG H, GUO L P, et al.Effect of Niobium Content on Irradiation Microstructure and Hardening in FeCrAl-Based Alloys[J]. Journal of Materials Science & Technology, 2021, 95: 181-192.
[112] DU D H, HUANG Y, LIN T H, et al.Elucidating the Role of Nb on the Long-Term Uniform Corrosion of FeCrAl Alloy in a Simulated PWR Environment[J]. Corrosion Science, 2025, 255: 113100.
[113] LIU Y, LUO B, HUANG H T, et al.Deuterium Permeation Behavior in a FeCrAl-Based Alloy Containing Mo, Nb and Ta Elements for LWR Cladding Application[J]. Journal of Nuclear Materials, 2022, 570: 153942.
[114] CHIANG H Y, WISS T, PARK S H, et al.TEM Analysis of Irradiation-Induced Interaction Layers in Coated UMo/X/Al Trilayer Systems (X=Ti, Nb, Zr, and Mo)[J]. Journal of Nuclear Materials, 2018, 499: 558-566.
[115] 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.