Investigation of stress corrosion properties of 9Cr-ODS-Al steel in supercritical water

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design Pub Date : 2025-03-24 DOI:10.1016/j.fusengdes.2025.114982

Rongrong Luo , Haiying Fu , Teng Zhang , Yongming Han , Yonghao Lu , Shilei Han , Hongfeng Zhu , Tengfei Yan , Pengyuan Li

{"title":"Investigation of stress corrosion properties of 9Cr-ODS-Al steel in supercritical water","authors":"Rongrong Luo , Haiying Fu , Teng Zhang , Yongming Han , Yonghao Lu , Shilei Han , Hongfeng Zhu , Tengfei Yan , Pengyuan Li","doi":"10.1016/j.fusengdes.2025.114982","DOIUrl":null,"url":null,"abstract":"<div><div>Consideration is being given to the Supercritical Water-Cooled Reactor (SCWR) as a prospective reactor configuration due to its superior thermal efficiency and the simplification of its structure, which obviates the need for steam generators and separators. This has led to the proposition of a supercritical water-based cooling concept within the scope of fusion reactor designs. When selecting alternative structural materials for the fusion breeding blanket, it is imperative to evaluate their resilience against corrosion and their vulnerability to stress corrosion cracking (SCC). In this context, the performance of 9Cr-ODS-Al steel, an oxide dispersion strengthened alloy, was examined under supercritical water (SCW) conditions at temperatures of 400 °C, 550 °C, and a pressure of 25 MPa over a period extending to 1152 h. The analytical findings from SEM and EDS confirmed the formation of a dual-layered oxide on the 9Cr-ODS-Al, consisting of an external layer and an internal compound oxide of Fe, Cr and Al. Slow Strain Rate Testing (SSRT) showed no signs of SCC on the examined fracture surfaces of the alloy, while a decrease in both ultimate tensile strength (UTS) and yield strength (YS) was observed at 550 °C/25MPa. Additionally, the fracture surfaces demonstrated characteristics of toughness, and the growth rate of stress corrosion cracks in the steel escalated with the stress intensity factor.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114982"},"PeriodicalIF":1.9000,"publicationDate":"2025-03-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":null,"platform":"Semanticscholar","paperid":null,"PeriodicalName":"Fusion Engineering and Design","FirstCategoryId":"5","ListUrlMain":"https://www.sciencedirect.com/science/article/pii/S0920379625001826","RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":null,"EPubDate":"","PubModel":"","JCR":"Q1","JCRName":"NUCLEAR SCIENCE & TECHNOLOGY","Score":null,"Total":0}

引用次数: 0

Abstract

Consideration is being given to the Supercritical Water-Cooled Reactor (SCWR) as a prospective reactor configuration due to its superior thermal efficiency and the simplification of its structure, which obviates the need for steam generators and separators. This has led to the proposition of a supercritical water-based cooling concept within the scope of fusion reactor designs. When selecting alternative structural materials for the fusion breeding blanket, it is imperative to evaluate their resilience against corrosion and their vulnerability to stress corrosion cracking (SCC). In this context, the performance of 9Cr-ODS-Al steel, an oxide dispersion strengthened alloy, was examined under supercritical water (SCW) conditions at temperatures of 400 °C, 550 °C, and a pressure of 25 MPa over a period extending to 1152 h. The analytical findings from SEM and EDS confirmed the formation of a dual-layered oxide on the 9Cr-ODS-Al, consisting of an external layer and an internal compound oxide of Fe, Cr and Al. Slow Strain Rate Testing (SSRT) showed no signs of SCC on the examined fracture surfaces of the alloy, while a decrease in both ultimate tensile strength (UTS) and yield strength (YS) was observed at 550 °C/25MPa. Additionally, the fracture surfaces demonstrated characteristics of toughness, and the growth rate of stress corrosion cracks in the steel escalated with the stress intensity factor.

查看原文本刊更多论文

由于超临界水冷反应堆（SCWR）热效率高，结构简单，无需蒸汽发生器和汽水分离器，因此正在考虑将其作为未来的反应堆结构。因此，在聚变反应堆设计范围内提出了超临界水冷却概念。在为聚变培养毯选择替代结构材料时，必须对其抗腐蚀能力和应力腐蚀开裂（SCC）的脆弱性进行评估。在此背景下，在超临界水（SCW）条件下，在温度为 400 °C、550 °C、压力为 25 兆帕的条件下，对 9Cr-ODS-Al 钢（一种氧化物分散强化合金）的性能进行了长达 1152 小时的检测。SEM 和 EDS 的分析结果证实，9Cr-ODS-Al 上形成了双层氧化物，由外部氧化层和内部铁、铬和铝的复合氧化物组成。慢应变速率测试（SSRT）显示，合金的受检断裂面上没有出现 SCC 的迹象，而在 550 °C/25MPa 下观察到极限拉伸强度（UTS）和屈服强度（YS）均有所下降。此外，断裂表面显示出韧性特征，钢中应力腐蚀裂纹的增长速度随应力强度因子的增加而增加。

本文章由计算机程序翻译，如有差异，请以英文原文为准。

求助全文

约1分钟内获得全文求助全文

来源期刊

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

CiteScore

3.50

自引率

23.50%

发文量

275

审稿时长

3.8 months

期刊介绍： The journal accepts papers about experiments (both plasma and technology), theory, models, methods, and designs in areas relating to technology, engineering, and applied science aspects of magnetic and inertial fusion energy. Specific areas of interest include: MFE and IFE design studies for experiments and reactors; fusion nuclear technologies and materials, including blankets and shields; analysis of reactor plasmas; plasma heating, fuelling, and vacuum systems; drivers, targets, and special technologies for IFE, controls and diagnostics; fuel cycle analysis and tritium reprocessing and handling; operations and remote maintenance of reactors; safety, decommissioning, and waste management; economic and environmental analysis of components and systems.