A comparative study of oxide scales formed on Alloy 690 in deaerated supercritical water and supercritical CO2 at 600 °C

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design Pub Date : 2025-09-29 DOI:10.1016/j.nucengdes.2025.114502

Jiamei Wang, Chen Peng, Kai Chen, Xianglong Guo, Lefu Zhang

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引用次数: 0

Abstract

Selecting an optimal material that offers a balanced combination of mechanical strength, outstanding corrosion resistance, and minimal neutron absorption remains a key challenge for both Generation-IV nuclear systems and materials science research. Austenitic alloys, including Fe-based stainless steels and Ni-based alloys, have emerged as promising alternatives to ferritic/martensitic (F/M) steels, owing to their superior corrosion resistance and improved creep performance. Among them, high-chromium Ni-based alloys demonstrate superior corrosion and oxidation resistance in high temperature steam-exceeding that of Fe-based austenitic stainless steels and F/M steels by over an order of magnitude. In this work, a comparative study of oxide scales formed on Alloy 690 in deaerated supercritical water and supercritical CO₂ at 600 °C was conducted. The study found that weight gains in both environments follow near-cubic rate laws. The superior oxidation resistance observed in both environments, compared to other pure austenitic alloys such as 800H, 316L, and 347, can mainly be attributed to its high chromium content. Nearly 1.5 times higher oxidation rate was observed in supercritical CO₂ than in SCW. A key observation was that the direct external oxidation and the rapid transition from internal to external oxidation within the initial 24-hours exposure in both two environments are responsible for its superior oxidation resistance. With prolonged exposure time, Cr-rich spinel oxides and Ni-rich networks within the internal oxidation zone gradually convert into Cr₂O₃, contributing to the growth of this protective chromia scale and thereby significantly retarding the oxidation process. Severe grain boundaries (GB) migration and oxidation was observed in both environments as the Cr-rich oxides at the GB was less effective at preventing oxidation, resulting in a significantly faster oxidation rate in these areas compared to the bulk grains. The slightly higher oxidation rate in supercritical CO₂ might be mainly attributed to the limited protection provided by the sparse outer NiO oxide scale and the breakdown of the innermost SiO₂ rich oxide film at O/M interface.

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690合金在脱氧超临界水和超临界CO2中600℃形成氧化鳞的比较研究

选择一种既能提供平衡的机械强度、优异的耐腐蚀性和最小中子吸收的最佳材料，仍然是第四代核系统和材料科学研究的关键挑战。奥氏体合金，包括铁基不锈钢和镍基合金，由于其优异的耐腐蚀性和改善的蠕变性能，已成为铁素体/马氏体（F/M）钢的有希望的替代品。其中，高铬镍基合金在高温蒸汽中表现出优异的耐腐蚀性和抗氧化性，比铁基奥氏体不锈钢和F/M钢高出一个数量级以上。本文对690合金在脱氧超临界水和超临界CO2中在600℃条件下形成的氧化鳞进行了对比研究。研究发现，在这两种环境下，体重增加都遵循近立方率定律。与其他纯奥氏体合金（如800H、316L和347）相比，在这两种环境下观察到的优越的抗氧化性主要归因于其高铬含量。超临界CO2处理的氧化速率是超临界水处理的近1.5倍。一个关键的观察结果是，在这两个环境中，在最初的24小时内，直接的外部氧化和从内部到外部氧化的快速过渡是其优越的抗氧化性的原因。随着暴露时间的延长，富cr尖晶石氧化物和内部氧化区内的富ni网络逐渐转化为Cr2O3，促进了保护色垢的生长，从而显著延缓了氧化过程。在两种环境下均观察到严重的晶界迁移和氧化，因为晶界处的富cr氧化物防止氧化的效果较差，导致这些区域的氧化速率明显快于大块晶粒。超临界CO2中略高的氧化速率可能主要是由于外层稀疏的氧化NiO氧化垢提供的保护有限，而最内层的富SiO2氧化膜在O/M界面处被击穿。

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

Nuclear Engineering and Design 工程技术-核科学技术

CiteScore

3.40

自引率

11.80%

发文量

377

审稿时长

5 months

期刊介绍： Nuclear Engineering and Design covers the wide range of disciplines involved in the engineering, design, safety and construction of nuclear fission reactors. The Editors welcome papers both on applied and innovative aspects and developments in nuclear science and technology. Fundamentals of Reactor Design include: • Thermal-Hydraulics and Core Physics • Safety Analysis, Risk Assessment (PSA) • Structural and Mechanical Engineering • Materials Science • Fuel Behavior and Design • Structural Plant Design • Engineering of Reactor Components • Experiments Aspects beyond fundamentals of Reactor Design covered: • Accident Mitigation Measures • Reactor Control Systems • Licensing Issues • Safeguard Engineering • Economy of Plants • Reprocessing / Waste Disposal • Applications of Nuclear Energy • Maintenance • Decommissioning Papers on new reactor ideas and developments (Generation IV reactors) such as inherently safe modular HTRs, High Performance LWRs/HWRs and LMFBs/GFR will be considered; Actinide Burners, Accelerator Driven Systems, Energy Amplifiers and other special designs of power and research reactors and their applications are also encouraged.