{"title":"ATR 辐照 HT9 包层 U-10M(10M = 5Mo-4.3Ti-0.7Zr wt%)金属燃料中燃料包层化学相互作用 (FCCI) 的透射电子显微镜特性分析","authors":"Yachun Wang, Jatuporn Burns, Tiankai Yao, Luca Capriotti","doi":"10.1016/j.jnucmat.2024.155209","DOIUrl":null,"url":null,"abstract":"<div><p>The pseudo-binary metallic fuel alloy, U-10M (wt%, M is the optimal combination of Mo, Ti, and Zr), has the potential to increase fuel solidus temperature, reduce the onset temperature of body-centered cubic phase, and increase the fuel's chemical stability compared with the conventional U-10Zr (wt%) metallic fuel. Post Irradiation Examination (PIE) confirmed excellent fuel performance for the U-10M (10M=5Mo-4.3Ti-0.7Zr wt%) fuel irradiated in the Advanced Test Reactor (ATR) to 2.2 at% burnup at Peak Inner Cladding Temperature (PICT) of 650 °C, an upper bound temperature for metallic fuel. But previous PIE study also observed Fuel Cladding Chemical Interaction (FCCI) on the cladding side, which is known as a fuel performance limiting issue but has not been fully understood yet. As an effort to improve the understanding of FCCI phenomenon, this study performed Scanning Electron Microscopy (SEM) and in-depth Transmission Electron Microscopy (TEM) characterization on a FCCI region. The examined FCCI region is dominated by (U, Zr)(Fe, Cr)<sub>2,</sub> suggesting that U-Fe interdiffusion reaction played a key role in inducing FCCI. Additionally, the FCCI boundary into the cladding consists of four distinctive phases, (U, Zr)(Fe, Cr)<sub>2</sub>, fcc-Cr, tetragonal UCr<sub>0.1</sub>Fe<sub>9.9</sub>Si<sub>2</sub>, intermetallic σ-FeCr, and lanthanide fission products at concentration up to ∼5.5 at%. Another goal of this study is to verify the involvement of Ti and Mo in FCCI formation on the cladding side. Observable Ti is found halfway of the thickness in the examined FCCI region, while 0.3–5 at% Mo is detected across the entire thickness of the examined FCCI region. Neither Ti nor Mo reacted with HT9 cladding constituents despite their diffusion footmark. Therefore, alloying Ti and Mo into the U-Zr fuel should not complicate the interdiffusion reaction on the HT9 cladding side.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":null,"pages":null},"PeriodicalIF":2.8000,"publicationDate":"2024-06-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":"{\"title\":\"Transmission electron microscopy characterization of Fuel Cladding Chemical Interaction (FCCI) in ATR-irradiated HT9 clad U-10M (10M = 5Mo-4.3Ti-0.7Zr wt%) metallic fuel\",\"authors\":\"Yachun Wang, Jatuporn Burns, Tiankai Yao, Luca Capriotti\",\"doi\":\"10.1016/j.jnucmat.2024.155209\",\"DOIUrl\":null,\"url\":null,\"abstract\":\"<div><p>The pseudo-binary metallic fuel alloy, U-10M (wt%, M is the optimal combination of Mo, Ti, and Zr), has the potential to increase fuel solidus temperature, reduce the onset temperature of body-centered cubic phase, and increase the fuel's chemical stability compared with the conventional U-10Zr (wt%) metallic fuel. Post Irradiation Examination (PIE) confirmed excellent fuel performance for the U-10M (10M=5Mo-4.3Ti-0.7Zr wt%) fuel irradiated in the Advanced Test Reactor (ATR) to 2.2 at% burnup at Peak Inner Cladding Temperature (PICT) of 650 °C, an upper bound temperature for metallic fuel. But previous PIE study also observed Fuel Cladding Chemical Interaction (FCCI) on the cladding side, which is known as a fuel performance limiting issue but has not been fully understood yet. As an effort to improve the understanding of FCCI phenomenon, this study performed Scanning Electron Microscopy (SEM) and in-depth Transmission Electron Microscopy (TEM) characterization on a FCCI region. The examined FCCI region is dominated by (U, Zr)(Fe, Cr)<sub>2,</sub> suggesting that U-Fe interdiffusion reaction played a key role in inducing FCCI. Additionally, the FCCI boundary into the cladding consists of four distinctive phases, (U, Zr)(Fe, Cr)<sub>2</sub>, fcc-Cr, tetragonal UCr<sub>0.1</sub>Fe<sub>9.9</sub>Si<sub>2</sub>, intermetallic σ-FeCr, and lanthanide fission products at concentration up to ∼5.5 at%. Another goal of this study is to verify the involvement of Ti and Mo in FCCI formation on the cladding side. Observable Ti is found halfway of the thickness in the examined FCCI region, while 0.3–5 at% Mo is detected across the entire thickness of the examined FCCI region. Neither Ti nor Mo reacted with HT9 cladding constituents despite their diffusion footmark. Therefore, alloying Ti and Mo into the U-Zr fuel should not complicate the interdiffusion reaction on the HT9 cladding side.</p></div>\",\"PeriodicalId\":373,\"journal\":{\"name\":\"Journal of Nuclear Materials\",\"volume\":null,\"pages\":null},\"PeriodicalIF\":2.8000,\"publicationDate\":\"2024-06-05\",\"publicationTypes\":\"Journal Article\",\"fieldsOfStudy\":null,\"isOpenAccess\":false,\"openAccessPdf\":\"\",\"citationCount\":\"0\",\"resultStr\":null,\"platform\":\"Semanticscholar\",\"paperid\":null,\"PeriodicalName\":\"Journal of Nuclear Materials\",\"FirstCategoryId\":\"5\",\"ListUrlMain\":\"https://www.sciencedirect.com/science/article/pii/S0022311524003118\",\"RegionNum\":2,\"RegionCategory\":\"工程技术\",\"ArticlePicture\":[],\"TitleCN\":null,\"AbstractTextCN\":null,\"PMCID\":null,\"EPubDate\":\"\",\"PubModel\":\"\",\"JCR\":\"Q3\",\"JCRName\":\"MATERIALS SCIENCE, MULTIDISCIPLINARY\",\"Score\":null,\"Total\":0}","platform":"Semanticscholar","paperid":null,"PeriodicalName":"Journal of Nuclear Materials","FirstCategoryId":"5","ListUrlMain":"https://www.sciencedirect.com/science/article/pii/S0022311524003118","RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":null,"EPubDate":"","PubModel":"","JCR":"Q3","JCRName":"MATERIALS SCIENCE, MULTIDISCIPLINARY","Score":null,"Total":0}
Transmission electron microscopy characterization of Fuel Cladding Chemical Interaction (FCCI) in ATR-irradiated HT9 clad U-10M (10M = 5Mo-4.3Ti-0.7Zr wt%) metallic fuel
The pseudo-binary metallic fuel alloy, U-10M (wt%, M is the optimal combination of Mo, Ti, and Zr), has the potential to increase fuel solidus temperature, reduce the onset temperature of body-centered cubic phase, and increase the fuel's chemical stability compared with the conventional U-10Zr (wt%) metallic fuel. Post Irradiation Examination (PIE) confirmed excellent fuel performance for the U-10M (10M=5Mo-4.3Ti-0.7Zr wt%) fuel irradiated in the Advanced Test Reactor (ATR) to 2.2 at% burnup at Peak Inner Cladding Temperature (PICT) of 650 °C, an upper bound temperature for metallic fuel. But previous PIE study also observed Fuel Cladding Chemical Interaction (FCCI) on the cladding side, which is known as a fuel performance limiting issue but has not been fully understood yet. As an effort to improve the understanding of FCCI phenomenon, this study performed Scanning Electron Microscopy (SEM) and in-depth Transmission Electron Microscopy (TEM) characterization on a FCCI region. The examined FCCI region is dominated by (U, Zr)(Fe, Cr)2, suggesting that U-Fe interdiffusion reaction played a key role in inducing FCCI. Additionally, the FCCI boundary into the cladding consists of four distinctive phases, (U, Zr)(Fe, Cr)2, fcc-Cr, tetragonal UCr0.1Fe9.9Si2, intermetallic σ-FeCr, and lanthanide fission products at concentration up to ∼5.5 at%. Another goal of this study is to verify the involvement of Ti and Mo in FCCI formation on the cladding side. Observable Ti is found halfway of the thickness in the examined FCCI region, while 0.3–5 at% Mo is detected across the entire thickness of the examined FCCI region. Neither Ti nor Mo reacted with HT9 cladding constituents despite their diffusion footmark. Therefore, alloying Ti and Mo into the U-Zr fuel should not complicate the interdiffusion reaction on the HT9 cladding side.
期刊介绍:
The Journal of Nuclear Materials publishes high quality papers in materials research for nuclear applications, primarily fission reactors, fusion reactors, and similar environments including radiation areas of charged particle accelerators. Both original research and critical review papers covering experimental, theoretical, and computational aspects of either fundamental or applied nature are welcome.
The breadth of the field is such that a wide range of processes and properties in the field of materials science and engineering is of interest to the readership, spanning atom-scale processes, microstructures, thermodynamics, mechanical properties, physical properties, and corrosion, for example.
Topics covered by JNM
Fission reactor materials, including fuels, cladding, core structures, pressure vessels, coolant interactions with materials, moderator and control components, fission product behavior.
Materials aspects of the entire fuel cycle.
Materials aspects of the actinides and their compounds.
Performance of nuclear waste materials; materials aspects of the immobilization of wastes.
Fusion reactor materials, including first walls, blankets, insulators and magnets.
Neutron and charged particle radiation effects in materials, including defects, transmutations, microstructures, phase changes and macroscopic properties.
Interaction of plasmas, ion beams, electron beams and electromagnetic radiation with materials relevant to nuclear systems.