通过评估Zr-Cr共晶对熔覆层结构完整性的影响,探索cr包覆ATF熔覆层的峰值熔覆温度极限

IF 3.2 2区 工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY Journal of Nuclear Materials Pub Date : 2025-02-01 Epub Date: 2024-12-18 DOI:10.1016/j.jnucmat.2024.155577
SungHoon Joung, Hyunwoo Yook, Dongju Kim, Youho Lee
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引用次数: 0

摘要

传统锆合金中不存在的Zr-Cr共晶(~ 1320°C)的形成,给cr涂层的事故容忍燃料(ATF)带来了重大的安全问题。本研究通过考察Zr-Cr共晶对cr包覆Zr-1.1Nb包覆层机械完整性的影响,研究了cr包覆ATF的峰值包覆温度(PCT)极限。为了实现这一目标,在蒸汽和无氧环境下对镀铬试样进行了冷却剂整体损失事故(LOCA)试验和环压缩试验(rct)。结果表明,虽然Zr和Cr之间共晶相的形成不会导致组织破坏,但会降低熔覆层的延展性。然而,在相同条件下,氧化的显著影响掩盖了Zr-Cr共晶对塑性降低的影响。在共晶起始温度以上氧化试样的严重延性损失的主要原因是在高温下氧扩散的增加。因此,与1204℃氧化试样相比,增加的氧浓度进一步降低了熔覆层的延展性。基于这些发现,在cr包覆ATF熔覆层中,由于Zr基体氧化导致的延展性明显降低,强调了只要cr包覆ATF熔覆层基体仍然是锆合金,就必须遵守当前PCT限值。此外,在2400°F(1315°C)以上的温度下,锆合金中观察到的过度脆化是建立当前2200°F(1204°C) PCT极限的关键因素。因此,考虑到在这些较高温度下氧气快速扩散和随之而来的延展性降低,将cr涂层ATF熔覆层的PCT极限延长到1204℃以上是不切实际的。因此,维持目前2200°F(1204°C)的PCT限制,可以作为在现有监管框架内确保cr涂层ATF包层安全的有效方法。
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Exploring the Peak Cladding Temperature Limit of Cr-coated ATF Cladding by Assessing the Impact of the Zr-Cr Eutectic on the Structural Integrity of Cladding
The formation of the Zr-Cr eutectic (∼1320 °C), which does not occur in conventional Zirconium alloys, introduces a significant safety concern for Cr-coated Accident Tolerant Fuel (ATF). This study investigated the Peak Cladding Temperature (PCT) limit for Cr-coated ATF by examining the effects of the Zr-Cr eutectic on the mechanical integrity of Cr-coated Zr-1.1Nb claddings. To achieve this, Integral Loss of Coolant Accident (LOCA) tests and Ring Compression Tests (RCTs) were conducted on Cr-coated specimens under both steam and oxygen-free environments. The results indicated that while the formation of the eutectic phase between Zr and Cr does not result in structural failure, it reduced the ductility of the cladding. However, the impact of Zr-Cr eutectic on the reduction in ductility was overshadowed by the significant impact of the oxidation under the same conditions. The primary cause of the severe ductility loss in specimens oxidized above the eutectic onset temperature was the increased oxygen diffusion at elevated temperatures. Consequently, compared to specimens oxidized at 1204 °C, the increased oxygen concentration in the ductile layer further reduced the ductility of the cladding. Based on these findings, the pronounced reduction in ductility caused by oxidation of the Zr matrix in Cr-coated ATF cladding underscored the necessity of adhering to the current PCT limit, as long as the cladding matrix of Cr-coated ATF cladding remains Zirconium alloy. Furthermore, the excessive embrittlement observed in Zirconium alloy at temperatures above 2400 °F (1315 °C) was a key factor in establishing the current 2200 °F (1204 °C) PCT limit. As a result, extending the PCT limit beyond 1204 °C for Cr-coated ATF cladding is impractical, given the rapid oxygen diffusion and the consequent reduction in ductility at these higher temperatures. Therefore, maintaining the current 2200 °F (1204 °C) PCT limit for Cr-coated ATF cladding can serve as the effective approach for ensuring the safety of Cr-coated ATF cladding within the existing regulatory framework.
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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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