FeCrAl fuel/clad chemical interaction in light water reactor environments

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY Journal of Nuclear Materials Pub Date : 2023-09-08 DOI:10.1016/j.jnucmat.2023.154717

Haozheng J. Qu , Maria Higgins , Hamdy Abouelella , Fabiola Cappia , Jatuporn Burns , Lingfeng He , Caleb Massey , Jason Harp , Kevin G. Field , Richard Howard , Rajnikant V. Umretiya , Andrew K. Hoffman , Janelle P. Wharry , Raul B. Rebak

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引用次数: 1

Abstract

This article investigates the fuel-cladding chemical interaction (FCCI) behavior of two commercial FeCrAl alloys, APMT composition (Fe-21Cr-5Al-3Mo wt.%) and C35M (Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.%), after neutron irradiation. “H-cup” diffusion multiples of FeCrAl alloys and ceramic UO₂ fuel were irradiated at a temperature of ∼300 °C to a total estimated burnup of 26 GWd/tHM. Post-irradiation Examination results demonstrate the excellent degradation resistance of FeCrAl alloys as accident tolerant fuel (ATF) cladding materials in light water reactor conditions. The study concludes that there was no irradiation-induced defects observed in either of the two commercial FeCrAl claddings. The formation of amorphous Al/U mixed oxide was observed at the fuel-clad interface, which can serve as a tritium permeation barrier and protect against potential chemical attack from the fuel. The study attributed the formation of amorphous Al/U mixed oxide to the low temperature and limited time of neutron irradiation. APMT forms more distinct Cr and Cr-Fe intermetallic at the FeCrAl-UO₂ interface than C35M due to the higher bulk Cr:Al ratio.

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轻水堆环境中FeCrAl燃料/包壳的化学相互作用

本文研究了两种商用FeCrAl合金APMT成分(Fe-21Cr-5Al-3Mo wt.%)和C35M成分(Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.%)在中子辐照后的燃料包层化学相互作用(FCCI)行为。“h杯”扩散倍的FeCrAl合金和陶瓷UO2燃料在~ 300°C的温度下辐照，估计总燃耗为26 GWd/tHM。辐照后试验结果表明，在轻水反应堆条件下，FeCrAl合金作为事故容忍燃料(ATF)包层材料具有优异的抗降解性能。该研究的结论是，在两种商用FeCrAl包层中均未观察到辐照引起的缺陷。在燃料包覆界面处形成无定形Al/U混合氧化物，可作为氚渗透屏障，防止燃料的潜在化学侵蚀。该研究将非晶态Al/U混合氧化物的形成归因于低温和有限的中子辐照时间。APMT在fecr - uo2界面形成的Cr和Cr- fe金属间化合物比C35M更明显，这是由于APMT具有更高的体积Cr:Al比。

本文章由计算机程序翻译，如有差异，请以英文原文为准。

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来源期刊

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.