研究HT9/U-10Zr金属快堆燃料中燃料包壳化学相互作用的综合方法:将机器学习与电子显微镜和局部机械性能分析相结合

IF 3.2 2区 工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY
Luca Capriotti, Fidelma Di Lemma, Daniele Salvato, Fei Xu, Yalei Tang, Kyle M. Paaren, Alexander L. Swearingen, Colby B. Jensen, Yachun Wang, Douglas L. Porter
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引用次数: 0

摘要

金属铀锆核燃料合金作为下一代钠冷快堆的有前途的候选材料,重新引起了人们的兴趣。最近的研究和技术评估已经确定了几个需要改进的领域,增加了知识和可靠的数据,以加强U-Zr燃料设计的资格和商业应用基础。影响该燃料系统性能的最具挑战性的现象之一是燃料包壳化学相互作用(FCCI)。本工作旨在通过综合方法检测原型全长燃料销中选定的HT9/U-10Zr (wt. %)燃料样品,以获取FCCI数据。该方法将扫描电镜(SEM)微观结构表征与局部力学性能检测相结合,以加深对HT9/U-10Zr燃料系统FCCI现象的理解。特别是,本研究侧重于在快通量试验设施(FFTF)辐照的MFF燃料销,旨在确定金属燃料作为FFTF驱动燃料的资格,并评估其在大型快堆中的可行性。电子显微镜为检测和区分不同的FCCI层提供了高置信度,而小规模力学测试(SSMT)则探测了这些层的力学性能。对MFF-2引脚192167进行扫描电镜检查,时间平均内包层温度(TICT)略高于500°C,显示最小的包层侧FCCI(包层损耗)。相比之下,在高温的MFF-3引脚193045和MFF-5引脚195011样品中观察到明显更厚的包层损耗,包括两个不同的子层,其中TICT范围为610-635°C。SSMT表明靠近燃料的亚层完全脆化,另一亚层有脆化的趋势。此外,一种新的机器学习方法被开发、验证并用于量化包层损耗厚度。机器学习方法可靠地预测了各种燃料销和样品横截面的损耗厚度。HT9/U-10Zr燃料系统的包层损耗数据表明,包层损耗具有很强的温度依赖性。然而,数据集仍然很小,持续的研究活动对于进一步了解FCCI现象和开发可靠的FCCI模型以增强各种条件下的燃料性能模拟至关重要。
本文章由计算机程序翻译,如有差异,请以英文原文为准。
An integrated approach to examine fuel-cladding chemical interaction in HT9/U-10Zr metallic fast reactor fuels: Coupling machine learning with electron microscopy and local mechanical properties analysis
The metallic U-Zr nuclear fuel alloy has garnered renewed interest as a promising candidate for next-generation sodium-cooled fast reactors. Recent studies and technology assessments have identified several areas requiring improvements, enhanced knowledge, and reliable data to strengthen the U-Zr fuel design basis for qualification and commercial applications. One of the most challenging phenomena impacting this fuel system’s performance is fuel-cladding chemical interaction (FCCI). This work aimed to harvest FCCI data by examining selected HT9/U-10Zr (wt. %) fuel samples of prototypic full-length fuel pins through an integrated approach. This approach integrated scanning electron microscopy (SEM) microstructure characterization with localized mechanical properties examination to deepen understanding of FCCI phenomenon in HT9/U-10Zr fuel system. Particularly, this study focused on MFF fuel pins irradiated at Fast Flux Test Facility (FFTF), which aimed to qualify metallic fuel as a driver fuel for FFTF and to assess its viability for larger-scale fast reactors. Electron microscopy provided high confidence in detecting and distinguishing the different FCCI layers, while small-scale mechanical testing (SSMT) probed the mechanical properties of these layers. SEM examination of a MFF-2 pin 192167, with a time averaged inner cladding temperature (TICT) slightly over 500°C, revealed minimal cladding-side FCCI (cladding wastage). In contrast, significantly thicker cladding wastage comprising two distinct sublayers was observed in samples from the thermally hot MFF-3 pin 193045 and MFF-5 pin 195011 where the TICT ranged from 610-635°C. SSMT indicated complete embrittlement in the sublayer adjacent to the fuel and a tendency toward embrittlement in the other sublayer. Additionally, a new machine learning method was developed, validated, and used to quantify cladding wastage thickness. The machine learning method reliably predicted the wastage thickness across various fuel pins and sample cross-sections. The available cladding wastage data from HT9/U-10Zr fuel system demonstrated a strong temperature dependency. However, the dataset remains small, and ongoing research activities are essential to further understand the FCCI phenomenon and develop a reliable FCCI model for enhanced fuel performance simulation under various conditions.
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来源期刊
Journal of Nuclear Materials
Journal of Nuclear Materials 工程技术-材料科学:综合
CiteScore
5.70
自引率
25.80%
发文量
601
审稿时长
63 days
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.
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