• 文献检索
  • 文档翻译
  • 深度研究
  • 学术资讯
  • Suppr Zotero 插件Zotero 插件
  • 邀请有礼
  • 套餐&价格
  • 历史记录
应用&插件
Suppr Zotero 插件Zotero 插件浏览器插件Mac 客户端Windows 客户端微信小程序
定价
高级版会员购买积分包购买API积分包
服务
文献检索文档翻译深度研究API 文档MCP 服务
关于我们
关于 Suppr公司介绍联系我们用户协议隐私条款
关注我们

Suppr 超能文献

核心技术专利：CN118964589B侵权必究
粤ICP备2023148730 号-1Suppr @ 2026

文献检索

告别复杂PubMed语法，用中文像聊天一样搜索，搜遍4000万医学文献。AI智能推荐，让科研检索更轻松。

立即免费搜索

文件翻译

保留排版，准确专业，支持PDF/Word/PPT等文件格式，支持 12+语言互译。

免费翻译文档

深度研究

AI帮你快速写综述，25分钟生成高质量综述，智能提取关键信息，辅助科研写作。

立即免费体验

超临界水环境中辐照12Cr铁素体-马氏体不锈钢应力腐蚀开裂抗性研究

Investigation of Stress Corrosion Cracking Resistance of Irradiated 12Cr Ferritic-Martensitic Stainless Steel in Supercritical Water Environment.

作者信息

Margolin Boris, Pirogova Natalia, Sorokin Alexander, Kokhonov Vasiliy, Dub Alexey, Safonov Ivan

机构信息

Central Research Institute of Structural Materials "Prometey" of National Research Center "Kurchatov Institute", 191015 Saint-Petersburg, Russia.

Joint-Stock Company "Science and Innovations", 115035 Moscow, Russia.

出版信息

Materials (Basel). 2023 Mar 24;16(7):2585. doi: 10.3390/ma16072585.

DOI:10.3390/ma16072585
PMID:37048889
原文链接:https://pmc.ncbi.nlm.nih.gov/articles/PMC10095323/
Abstract

The supercritical water-cooled reactors (SWCR) belong to Generation IV of reactors. These reactors have a number of advantages over currently operating WWERs and PWRs. These advantages include higher thermal efficiency, a more simplified unit design, and the possibility of incorporating it into a closed fuel cycle. It is therefore necessary to identify candidate materials for the SWCR and validate the safety and effectiveness of their use. 12Cr ferritic-martensitic (F/M) stainless steel is considered a candidate material for SWCR internals. Radiation embrittlement and corrosion cracking in the primary circuit coolant environment are the main mechanisms of F/M steels degradation during SWCR operation. Here, the stress corrosion cracking (SCC) in supercritical water at 390 and 550 °C of 12Cr F/M steel irradiated by neutrons to 12 dpa is investigated. Autoclave tests of specially designed disk specimens in supercritical water were performed. The tests were carried out under different constant load (CL), temperature 450 °C, and pressure in autoclave 25 MPa. The threshold stress, below which the SCC initiation of irradiated 12Cr F/M steel does not occur, was determined.

摘要

超临界水冷反应堆（SWCR）属于第四代反应堆。与目前运行的压水堆和沸水堆相比，这些反应堆具有许多优势。这些优势包括更高的热效率、更简化的机组设计以及将其纳入闭式燃料循环的可能性。因此，有必要确定超临界水冷反应堆的候选材料，并验证其使用的安全性和有效性。12Cr铁素体-马氏体（F/M）不锈钢被认为是超临界水冷反应堆内部构件的候选材料。在超临界水冷反应堆运行期间，主回路冷却剂环境中的辐射脆化和腐蚀开裂是F/M钢降解的主要机制。在此，研究了中子辐照至12 dpa的12Cr F/M钢在390和550°C超临界水中的应力腐蚀开裂（SCC）情况。对专门设计的圆盘试样在超临界水中进行了高压釜试验。试验在不同的恒定载荷（CL）、450°C温度和高压釜压力25 MPa条件下进行。确定了辐照12Cr F/M钢不发生应力腐蚀开裂起始的阈值应力。

https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/07a018e7e769/materials-16-02585-g025.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/f61ac7b578f3/materials-16-02585-g001.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/237d4e48a839/materials-16-02585-g002.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/4ea0781d8d00/materials-16-02585-g003.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/ef21821ff178/materials-16-02585-g004.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/55f1167117d6/materials-16-02585-g005.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/5bd455c46775/materials-16-02585-g006.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/28ac847b5279/materials-16-02585-g007a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/e93e84766930/materials-16-02585-g008a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/50f3f50b3e9f/materials-16-02585-g009a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/9ebb78091322/materials-16-02585-g010a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/a864702091a4/materials-16-02585-g011a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/380371950739/materials-16-02585-g012a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/3b05f40787b6/materials-16-02585-g013.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/923c087d1e31/materials-16-02585-g014.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/f8147ec823f6/materials-16-02585-g015.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/9d3607faec0b/materials-16-02585-g016.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/3fe7d6d88e90/materials-16-02585-g017.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/2ae6bda16358/materials-16-02585-g018.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/5b3e9057ebc2/materials-16-02585-g019.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/52544cd648f9/materials-16-02585-g020.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/6ff35b86c2f6/materials-16-02585-g021a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/619b230dcbc4/materials-16-02585-g022.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/6a2762be39f6/materials-16-02585-g023.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/e3bbcf6a0a3f/materials-16-02585-g024a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/07a018e7e769/materials-16-02585-g025.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/f61ac7b578f3/materials-16-02585-g001.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/237d4e48a839/materials-16-02585-g002.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/4ea0781d8d00/materials-16-02585-g003.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/ef21821ff178/materials-16-02585-g004.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/55f1167117d6/materials-16-02585-g005.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/5bd455c46775/materials-16-02585-g006.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/28ac847b5279/materials-16-02585-g007a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/e93e84766930/materials-16-02585-g008a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/50f3f50b3e9f/materials-16-02585-g009a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/9ebb78091322/materials-16-02585-g010a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/a864702091a4/materials-16-02585-g011a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/380371950739/materials-16-02585-g012a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/3b05f40787b6/materials-16-02585-g013.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/923c087d1e31/materials-16-02585-g014.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/f8147ec823f6/materials-16-02585-g015.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/9d3607faec0b/materials-16-02585-g016.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/3fe7d6d88e90/materials-16-02585-g017.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/2ae6bda16358/materials-16-02585-g018.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/5b3e9057ebc2/materials-16-02585-g019.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/52544cd648f9/materials-16-02585-g020.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/6ff35b86c2f6/materials-16-02585-g021a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/619b230dcbc4/materials-16-02585-g022.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/6a2762be39f6/materials-16-02585-g023.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/e3bbcf6a0a3f/materials-16-02585-g024a.jpg
https://cdn.ncbi.nlm.nih.gov/pmc/blobs/3b6f/10095323/07a018e7e769/materials-16-02585-g025.jpg

相似文献

1
Investigation of Stress Corrosion Cracking Resistance of Irradiated 12Cr Ferritic-Martensitic Stainless Steel in Supercritical Water Environment.超临界水环境中辐照12Cr铁素体-马氏体不锈钢应力腐蚀开裂抗性研究
Materials (Basel). 2023 Mar 24;16(7):2585. doi: 10.3390/ma16072585.
2
Corrosion Behavior of Ferritic 12Cr ODS and Martensitic X46Cr13 Steels in Nitric Acid and Sodium Chloride Solutions.铁素体12Cr氧化物弥散强化钢和马氏体X46Cr13钢在硝酸和氯化钠溶液中的腐蚀行为
Materials (Basel). 2024 Jul 12;17(14):3466. doi: 10.3390/ma17143466.
3
The Irradiation Effects in Ferritic, Ferritic-Martensitic and Austenitic Oxide Dispersion Strengthened Alloys: A Review.铁素体、铁素体-马氏体和奥氏体氧化物弥散强化合金中的辐照效应综述
Materials (Basel). 2024 Jul 10;17(14):3409. doi: 10.3390/ma17143409.
4
Effect of Applied Stress on T91 Steel Performance in Liquid Lead at 400 °C.外加应力对T91钢在400℃液态铅中性能的影响。
Materials (Basel). 2018 Dec 11;11(12):2512. doi: 10.3390/ma11122512.
5
The Effect of Holding Time on Dissimilar Transient Liquid-Phase-Bonded Properties of Super-Ferritic Stainless Steel 446 to Martensitic Stainless Steel 410 Using a Nickel-Based Interlayer.保温时间对采用镍基中间层实现超级铁素体不锈钢446与马氏体不锈钢410的异种瞬态液相连接性能的影响
Micromachines (Basel). 2022 Oct 22;13(11):1801. doi: 10.3390/mi13111801.
6
Influence of Chloride Concentration on Stress Corrosion Cracking and Crevice Corrosion of Austenitic Stainless Steel in Saline Environments.氯化物浓度对奥氏体不锈钢在盐环境中应力腐蚀开裂和缝隙腐蚀的影响。
Materials (Basel). 2020 Dec 10;13(24):5640. doi: 10.3390/ma13245640.
7
Electrochemical and passivation behavior investigation of ferritic stainless steel in simulated concrete pore media.铁素体不锈钢在模拟混凝土孔隙介质中的电化学与钝化行为研究
Data Brief. 2015 Sep 1;5:171-8. doi: 10.1016/j.dib.2015.08.016. eCollection 2015 Dec.
8
Improved Mechanical and Corrosion Properties of Powder Metallurgy Austenitic, Ferritic, and Martensitic Stainless Steels by Liquid Phase Sintering.通过液相烧结改善粉末冶金奥氏体、铁素体和马氏体不锈钢的力学性能和耐腐蚀性。
Materials (Basel). 2022 Aug 9;15(16):5483. doi: 10.3390/ma15165483.
9
A Comparative Study of Localized Corrosion and Stress Corrosion Cracking of 13Cr Martensitic Stainless Steel Using Acoustic Emission and X-ray Computed Tomography.使用声发射和X射线计算机断层扫描对13Cr马氏体不锈钢的局部腐蚀和应力腐蚀开裂的对比研究
Materials (Basel). 2019 Aug 12;12(16):2569. doi: 10.3390/ma12162569.
10
Modelling and Analysis of the Corrosion Characteristics of Ferritic-Martensitic Steels in Supercritical Water.铁素体-马氏体钢在超临界水中的腐蚀特性建模与分析
Materials (Basel). 2019 Jan 28;12(3):409. doi: 10.3390/ma12030409.