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用于核燃料包壳的新一代铁铬铝合金的水热腐蚀

Hydrothermal Corrosion of Latest Generation of FeCrAl Alloys for Nuclear Fuel Cladding.

作者信息

Nagothi Bhavani Sasank, Qu Haozheng, Zhang Wanming, Umretiya Rajnikant V, Dolley Evan, Rebak Raul B

机构信息

General Electric Research Center, 1 Research Circle, Schenectady, NY 12309, USA.

College of Nanotechnology, Science, and Engineering, University at Albany, Albany, NY 12203, USA.

出版信息

Materials (Basel). 2024 Apr 3;17(7):1633. doi: 10.3390/ma17071633.

DOI:10.3390/ma17071633
PMID:38612147
原文链接:https://pmc.ncbi.nlm.nih.gov/articles/PMC11012897/
Abstract

After the Fukushima nuclear disaster, the nuclear materials community has been vastly investing in accident tolerant fuel (ATF) concepts to modify/replace Zircaloy cladding material. Iron-chromium-aluminum (FeCrAl) alloys are one of the leading contenders in this race. In this study, we investigated FA-SMT (or APMT-2), PM-C26M, and Fe17Cr5.5Al over a time period of 6 months in simulated BWR environments and compared their performance with standard Zirc-2 and SS316 materials. Our results implied that water chemistry along with alloy chemistry has a profound effect on the corrosion rate of FeCrAl alloys. Apart from SS316 and Zirc-2 tube specimens, all FeCrAl alloys showed a mass loss in hydrogen water chemistry (HWC). FA-SMT displayed minimal mass loss compared to PM-C26M and Fe17Cr5.5Al because of its higher Cr content. The mass gain of FeCrAl alloys in normal water chemistry (NWC) is significantly less when compared to Zirc-2.

摘要

福岛核灾难发生后，核材料领域一直在大力投入研发事故容错燃料（ATF）概念，以改进/替代锆合金包壳材料。铁铬铝合金（FeCrAl）是这场竞争中的主要竞争者之一。在本研究中，我们在模拟沸水反应堆（BWR）环境中对FA-SMT（或APMT-2）、PM-C26M和Fe17Cr5.5Al进行了为期6个月的研究，并将它们的性能与标准锆-2和SS316材料进行了比较。我们的结果表明，水化学以及合金化学对FeCrAl合金的腐蚀速率有深远影响。除了SS316和锆-2管试样外，所有FeCrAl合金在氢水化学（HWC）中都出现了质量损失。由于FA-SMT的铬含量较高，与PM-C26M和Fe17Cr5.5Al相比，其质量损失最小。与锆-2相比，FeCrAl合金在正常水化学（NWC）中的质量增加明显较少。

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