{"title":"辐照U-10Mo杨氏模量的退化","authors":"Chaoyue Jin , Zhexiao Xie , Luning Chen , Xingdi Chen , Jing Zhang , Shurong Ding , Xiaobin Jian","doi":"10.1016/j.jnucmat.2025.155799","DOIUrl":null,"url":null,"abstract":"<div><div>Theoretical analysis of the four-point bending experimental results from the reference has demonstrated that the effective Young's modulus of heavily-irradiated U-10Mo fuel undergoes a significant reduction. However, the underlying mechanisms need to be fully elucidated. In this study, the irradiation-induced thermo-mechanical coupling behaviors of monolithic fuel plates are first numerically investigated by employing the fuel skeleton creep-based volumetric growth strain model and the porosity-related macroscale creep rate model for the contained U-10Mo fuel foils. The predicted average thicknesses for the bending specimens from several fuel plates align well with the experimental measurements, validating the adopted models, algorithms and the obtained macroscale porosity values for irradiated U-10Mo fuel. The values of effective Young's modulus of U-10Mo fuel after different levels of irradiation are identified through the subsequent direct simulations of the four-point bending tests, with the numerically acquired macroscale mechanical responses of irradiated U-10Mo specimens matching the experimental data. After eliminating the effects of fuel porosity, it is found that the values of Young's modulus of dense U-10Mo fuel skeleton decrease with increasing fission density or macroscale porosity, thereby becoming a primary contributor to the degradation of the effective Young's modulus of irradiated U-10Mo fuel. Furthermore, mathematical models for the Young's modulus of irradiated U-10Mo fuel skeleton at room temperature are developed as functions of fission density and macroscale porosity, respectively. The predicted results indicate that the von Mises stress will significantly decrease and the equivalent creep strains might have a distinct increase, when the degradation of Young's modulus of fuel skeleton is incorporated. This work provides a foundation for the high-precise modeling of the irradiation-induced thermo-mechanical behaviors of the U-10Mo-based fuel elements or assemblies.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155799"},"PeriodicalIF":2.8000,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":"{\"title\":\"On the degradation of Young's modulus of irradiated U-10Mo\",\"authors\":\"Chaoyue Jin , Zhexiao Xie , Luning Chen , Xingdi Chen , Jing Zhang , Shurong Ding , Xiaobin Jian\",\"doi\":\"10.1016/j.jnucmat.2025.155799\",\"DOIUrl\":null,\"url\":null,\"abstract\":\"<div><div>Theoretical analysis of the four-point bending experimental results from the reference has demonstrated that the effective Young's modulus of heavily-irradiated U-10Mo fuel undergoes a significant reduction. However, the underlying mechanisms need to be fully elucidated. In this study, the irradiation-induced thermo-mechanical coupling behaviors of monolithic fuel plates are first numerically investigated by employing the fuel skeleton creep-based volumetric growth strain model and the porosity-related macroscale creep rate model for the contained U-10Mo fuel foils. The predicted average thicknesses for the bending specimens from several fuel plates align well with the experimental measurements, validating the adopted models, algorithms and the obtained macroscale porosity values for irradiated U-10Mo fuel. The values of effective Young's modulus of U-10Mo fuel after different levels of irradiation are identified through the subsequent direct simulations of the four-point bending tests, with the numerically acquired macroscale mechanical responses of irradiated U-10Mo specimens matching the experimental data. After eliminating the effects of fuel porosity, it is found that the values of Young's modulus of dense U-10Mo fuel skeleton decrease with increasing fission density or macroscale porosity, thereby becoming a primary contributor to the degradation of the effective Young's modulus of irradiated U-10Mo fuel. Furthermore, mathematical models for the Young's modulus of irradiated U-10Mo fuel skeleton at room temperature are developed as functions of fission density and macroscale porosity, respectively. The predicted results indicate that the von Mises stress will significantly decrease and the equivalent creep strains might have a distinct increase, when the degradation of Young's modulus of fuel skeleton is incorporated. This work provides a foundation for the high-precise modeling of the irradiation-induced thermo-mechanical behaviors of the U-10Mo-based fuel elements or assemblies.</div></div>\",\"PeriodicalId\":373,\"journal\":{\"name\":\"Journal of Nuclear Materials\",\"volume\":\"610 \",\"pages\":\"Article 155799\"},\"PeriodicalIF\":2.8000,\"publicationDate\":\"2025-04-01\",\"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/S0022311525001941\",\"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/S0022311525001941","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}
On the degradation of Young's modulus of irradiated U-10Mo
Theoretical analysis of the four-point bending experimental results from the reference has demonstrated that the effective Young's modulus of heavily-irradiated U-10Mo fuel undergoes a significant reduction. However, the underlying mechanisms need to be fully elucidated. In this study, the irradiation-induced thermo-mechanical coupling behaviors of monolithic fuel plates are first numerically investigated by employing the fuel skeleton creep-based volumetric growth strain model and the porosity-related macroscale creep rate model for the contained U-10Mo fuel foils. The predicted average thicknesses for the bending specimens from several fuel plates align well with the experimental measurements, validating the adopted models, algorithms and the obtained macroscale porosity values for irradiated U-10Mo fuel. The values of effective Young's modulus of U-10Mo fuel after different levels of irradiation are identified through the subsequent direct simulations of the four-point bending tests, with the numerically acquired macroscale mechanical responses of irradiated U-10Mo specimens matching the experimental data. After eliminating the effects of fuel porosity, it is found that the values of Young's modulus of dense U-10Mo fuel skeleton decrease with increasing fission density or macroscale porosity, thereby becoming a primary contributor to the degradation of the effective Young's modulus of irradiated U-10Mo fuel. Furthermore, mathematical models for the Young's modulus of irradiated U-10Mo fuel skeleton at room temperature are developed as functions of fission density and macroscale porosity, respectively. The predicted results indicate that the von Mises stress will significantly decrease and the equivalent creep strains might have a distinct increase, when the degradation of Young's modulus of fuel skeleton is incorporated. This work provides a foundation for the high-precise modeling of the irradiation-induced thermo-mechanical behaviors of the U-10Mo-based fuel elements or assemblies.
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.