Semi-integral LOCA testing of Cr-coated Optimized ZIRLOTM claddings

IF 3.2 2区 工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY Journal of Nuclear Materials Pub Date : 2025-03-18 DOI:10.1016/j.jnucmat.2025.155766
Ioannis Alakiozidis , Marc Lopes Nunes de Sousa , Axel Gauthier , Callum Hunt , Mia Maric , Antoine Ambard , Zaheen Shah , Philipp Frankel
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Abstract

Chromium (Cr)-coatings on zirconium-(Zr) based claddings have emerged as a promising short-term solution to enhance the accident tolerance of fuel assemblies in pressurised water reactors (PWRs) during loss-of-coolant accidents (LOCAs). In this study, we tested a large number (36 rods in total, each 30cm long) of uncoated and Cr-coated Optimized ZIRLOTM claddings under thermomechanical conditions that closely resemble a real LOCA. A unique experimental apparatus was employed to integrate multiple LOCA effects into a single test sequence, enabling a more accurate prediction of the performance of Cr-coatings and degradation mechanisms of the coated claddings. More specifically, the test sequence included: i) thermal ramping from 350–1200°C under varying internal pressures and heating rates in flowing steam; ii) isothermal steam oxidation at 1200°C for different durations; ii) cooling to 700°C followed by water quenching to 135°C; iv) partial-axial constraint at 135°C with load hold of 540N for 20s. Various characterisation techniques, including optical and scanning electron microscopy (SEM), 3D laser scanning, electron backscattered diffraction (EBSD), hardness testing and hydrogen analysis, were used to characterise the post-LOCA cladding microstructures. We found that Cr-coatings increased the burst temperature of uncoated claddings by ∼ 25–150°C and reduced the strain-to-burst and cladding deformation within 20 mm away from the burst opening. The magnitude of these improvements depended on the initial testing conditions and were more pronounced for the helium-propelled cold spray (HCS) coating, while less pronounced for the nitrogen-propelled CS (NCS) and physical vapour deposition (PVD) coatings. Additionally, we found that Cr-coatings increased the time threshold before significant cladding embrittlement by ∼100–555s compared to uncoated claddings. Finally, we concluded that when multiple LOCA effects are considered, predictions of additional coping time during a LOCA provided by the Cr-coatings are more conservative compared to single-factor tests.
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cr包覆优化ZIRLOTM包层的半积分LOCA测试
锆(Zr)基覆层上的铬(Cr)涂层已成为一种很有前途的短期解决方案,可提高压水堆(PWR)燃料组件在失冷事故(LOCA)期间的事故耐受性。在这项研究中,我们在与真实 LOCA 非常相似的热机械条件下测试了大量(共 36 根棒,每根 30 厘米长)无涂层和有铬涂层的优化 ZIRLOTM 包壳。采用了一种独特的实验装置,将多种 LOCA 效应整合到一个测试序列中,从而能够更准确地预测铬涂层的性能和涂层包层的降解机制。更具体地说,测试序列包括:i) 在流动蒸汽中以不同的内部压力和加热速率从 350°C 升温到 1200°C;ii) 在 1200°C 进行不同持续时间的等温蒸汽氧化;ii) 冷却到 700°C,然后水淬到 135°C;iv) 在 135°C 进行部分轴向约束,负载保持 540N 20 秒。我们采用了各种表征技术,包括光学和扫描电子显微镜 (SEM)、三维激光扫描、电子背散射衍射 (EBSD)、硬度测试和氢分析,来表征 LOCA 后包层的微观结构。我们发现,Cr 涂层将未涂层包层的爆裂温度提高了 25-150°C 左右,并降低了爆裂应变和距爆裂口 20 毫米范围内的包层变形。这些改善的幅度取决于初始测试条件,氦气推进冷喷(HCS)涂层的改善幅度更大,而氮气推进 CS(NCS)和物理气相沉积(PVD)涂层的改善幅度较小。此外,我们还发现,与未涂层的覆层相比,Cr 涂层将覆层发生明显脆化之前的时间阈值提高了 100-555 秒。最后,我们得出结论,当考虑到多重 LOCA 影响时,与单因素测试相比,Cr 涂层提供的 LOCA 期间额外应对时间的预测更为保守。
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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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