Chundong Wang , Xijie Wu , Jie Pan , Huawei Zhang , Jun Li , Jiwei Lin , Ping Cao , Cheng Su , Cong Li , Meiyi Yao , Xueshan Xiao
{"title":"FeCrAlMo-xZr燃料包层合金热腐蚀过程的显微组织与力学演化","authors":"Chundong Wang , Xijie Wu , Jie Pan , Huawei Zhang , Jun Li , Jiwei Lin , Ping Cao , Cheng Su , Cong Li , Meiyi Yao , Xueshan Xiao","doi":"10.1016/j.net.2025.103637","DOIUrl":null,"url":null,"abstract":"<div><div>The microstructural and mechanical evolution of the Fe-13Cr-6Al-2Mo-<em>x</em>Zr (<em>x</em> = 0.15, 0.5, 1.0 wt%) alloys for fuel cladding throughout various hot corrosion processes, such as thermal aging at 475 °C in air and high temperature oxidation at 1200 °C under an atmosphere of Ar + water vapour, were investigated. The findings indicate that as the Zr content increases, the grain size of the alloys becomes progressively smaller. During thermal aging, the secondary Laves phase precipitated within the grains and Mo segregation occurred on the grain boundaries, ultimately forming the μ phase. The plasticity decreased by degrees with aging time due to the formation and growth of precipitates, while the tensile strength increased only during the early aging period and then remained almost unchanged. The steam oxidation rate was promoted with the Zr content because of the smaller grain size and the deeper internal oxidation resulting from the faster diffusion of oxygen in ZrO<sub>2</sub> than that in Al<sub>2</sub>O<sub>3</sub>.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 9","pages":"Article 103637"},"PeriodicalIF":2.6000,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":"{\"title\":\"Microstructure and mechanical evolution of FeCrAlMo-xZr fuel cladding alloys during hot corrosion\",\"authors\":\"Chundong Wang , Xijie Wu , Jie Pan , Huawei Zhang , Jun Li , Jiwei Lin , Ping Cao , Cheng Su , Cong Li , Meiyi Yao , Xueshan Xiao\",\"doi\":\"10.1016/j.net.2025.103637\",\"DOIUrl\":null,\"url\":null,\"abstract\":\"<div><div>The microstructural and mechanical evolution of the Fe-13Cr-6Al-2Mo-<em>x</em>Zr (<em>x</em> = 0.15, 0.5, 1.0 wt%) alloys for fuel cladding throughout various hot corrosion processes, such as thermal aging at 475 °C in air and high temperature oxidation at 1200 °C under an atmosphere of Ar + water vapour, were investigated. The findings indicate that as the Zr content increases, the grain size of the alloys becomes progressively smaller. During thermal aging, the secondary Laves phase precipitated within the grains and Mo segregation occurred on the grain boundaries, ultimately forming the μ phase. The plasticity decreased by degrees with aging time due to the formation and growth of precipitates, while the tensile strength increased only during the early aging period and then remained almost unchanged. The steam oxidation rate was promoted with the Zr content because of the smaller grain size and the deeper internal oxidation resulting from the faster diffusion of oxygen in ZrO<sub>2</sub> than that in Al<sub>2</sub>O<sub>3</sub>.</div></div>\",\"PeriodicalId\":19272,\"journal\":{\"name\":\"Nuclear Engineering and Technology\",\"volume\":\"57 9\",\"pages\":\"Article 103637\"},\"PeriodicalIF\":2.6000,\"publicationDate\":\"2025-04-09\",\"publicationTypes\":\"Journal Article\",\"fieldsOfStudy\":null,\"isOpenAccess\":false,\"openAccessPdf\":\"\",\"citationCount\":\"0\",\"resultStr\":null,\"platform\":\"Semanticscholar\",\"paperid\":null,\"PeriodicalName\":\"Nuclear Engineering and Technology\",\"FirstCategoryId\":\"5\",\"ListUrlMain\":\"https://www.sciencedirect.com/science/article/pii/S1738573325002050\",\"RegionNum\":3,\"RegionCategory\":\"工程技术\",\"ArticlePicture\":[],\"TitleCN\":null,\"AbstractTextCN\":null,\"PMCID\":null,\"EPubDate\":\"\",\"PubModel\":\"\",\"JCR\":\"Q1\",\"JCRName\":\"NUCLEAR SCIENCE & TECHNOLOGY\",\"Score\":null,\"Total\":0}","platform":"Semanticscholar","paperid":null,"PeriodicalName":"Nuclear Engineering and Technology","FirstCategoryId":"5","ListUrlMain":"https://www.sciencedirect.com/science/article/pii/S1738573325002050","RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":null,"EPubDate":"","PubModel":"","JCR":"Q1","JCRName":"NUCLEAR SCIENCE & TECHNOLOGY","Score":null,"Total":0}
Microstructure and mechanical evolution of FeCrAlMo-xZr fuel cladding alloys during hot corrosion
The microstructural and mechanical evolution of the Fe-13Cr-6Al-2Mo-xZr (x = 0.15, 0.5, 1.0 wt%) alloys for fuel cladding throughout various hot corrosion processes, such as thermal aging at 475 °C in air and high temperature oxidation at 1200 °C under an atmosphere of Ar + water vapour, were investigated. The findings indicate that as the Zr content increases, the grain size of the alloys becomes progressively smaller. During thermal aging, the secondary Laves phase precipitated within the grains and Mo segregation occurred on the grain boundaries, ultimately forming the μ phase. The plasticity decreased by degrees with aging time due to the formation and growth of precipitates, while the tensile strength increased only during the early aging period and then remained almost unchanged. The steam oxidation rate was promoted with the Zr content because of the smaller grain size and the deeper internal oxidation resulting from the faster diffusion of oxygen in ZrO2 than that in Al2O3.
期刊介绍:
Nuclear Engineering and Technology (NET), an international journal of the Korean Nuclear Society (KNS), publishes peer-reviewed papers on original research, ideas and developments in all areas of the field of nuclear science and technology. NET bimonthly publishes original articles, reviews, and technical notes. The journal is listed in the Science Citation Index Expanded (SCIE) of Thomson Reuters.
NET covers all fields for peaceful utilization of nuclear energy and radiation as follows:
1) Reactor Physics
2) Thermal Hydraulics
3) Nuclear Safety
4) Nuclear I&C
5) Nuclear Physics, Fusion, and Laser Technology
6) Nuclear Fuel Cycle and Radioactive Waste Management
7) Nuclear Fuel and Reactor Materials
8) Radiation Application
9) Radiation Protection
10) Nuclear Structural Analysis and Plant Management & Maintenance
11) Nuclear Policy, Economics, and Human Resource Development