Impact of composition and cold work on SCC of high Si-bearing stainless steel simulating radiation-induced segregation under simulated PWR primary water

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY Journal of Nuclear Materials Pub Date : 2025-05-01 Epub Date: 2025-03-22 DOI:10.1016/j.jnucmat.2025.155775

Caibo Xie , Deng Pan , Hao Liu , Zhanpeng Lu , Tongming Cui , Junjie Chen , Xinhe Xu

{"title":"Impact of composition and cold work on SCC of high Si-bearing stainless steel simulating radiation-induced segregation under simulated PWR primary water","authors":"Caibo Xie , Deng Pan , Hao Liu , Zhanpeng Lu , Tongming Cui , Junjie Chen , Xinhe Xu","doi":"10.1016/j.jnucmat.2025.155775","DOIUrl":null,"url":null,"abstract":"<div><div>Thick block alloys simulating radiation-induced segregation effects of compositions at grain boundaries were designed and prepared, and deformation due to irradiation hardening was simulated by cold working treatments. The stress corrosion cracking (SCC) behavior of 17Cr-13Ni-3Si, 13Cr-17Ni, and 13Cr-17Ni-3Si alloys in the as-cast and cold-worked states was studied by crack growth rate (CGR) tests using compact tension specimens in hydrogenated high-temperature water. The fracture surfaces of the three as-cast state simulated alloys showed local SCC indications, while the fracture surfaces of all cold-worked state simulated alloys showed typical intergranular SCC characteristics. In the absence of cold work, the single Si addition has no noticeable effect on SCC propagation. Stress-accelerated grain boundary oxidation acts as the initiation site for intergranular cracks. Cold-worked low Cr alloy with high Si content significantly promotes SCC growth in hydrogenated water. The effect of alloy composition and strain hardening on SCC and its implications for irradiation-assisted SCC were discussed.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155775"},"PeriodicalIF":3.2000,"publicationDate":"2025-05-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":null,"platform":"Semanticscholar","paperid":null,"PeriodicalName":"Journal of Nuclear Materials","FirstCategoryId":"5","ListUrlMain":"https://www.sciencedirect.com/science/article/pii/S0022311525001709","RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":null,"EPubDate":"2025/3/22 0:00:00","PubModel":"Epub","JCR":"Q3","JCRName":"MATERIALS SCIENCE, MULTIDISCIPLINARY","Score":null,"Total":0}

引用次数: 0

Abstract

Thick block alloys simulating radiation-induced segregation effects of compositions at grain boundaries were designed and prepared, and deformation due to irradiation hardening was simulated by cold working treatments. The stress corrosion cracking (SCC) behavior of 17Cr-13Ni-3Si, 13Cr-17Ni, and 13Cr-17Ni-3Si alloys in the as-cast and cold-worked states was studied by crack growth rate (CGR) tests using compact tension specimens in hydrogenated high-temperature water. The fracture surfaces of the three as-cast state simulated alloys showed local SCC indications, while the fracture surfaces of all cold-worked state simulated alloys showed typical intergranular SCC characteristics. In the absence of cold work, the single Si addition has no noticeable effect on SCC propagation. Stress-accelerated grain boundary oxidation acts as the initiation site for intergranular cracks. Cold-worked low Cr alloy with high Si content significantly promotes SCC growth in hydrogenated water. The effect of alloy composition and strain hardening on SCC and its implications for irradiation-assisted SCC were discussed.

查看原文

微信好友朋友圈 QQ好友复制链接

本刊更多论文

成分和冷加工对高硅不锈钢模拟压水堆一次水辐射偏析SCC的影响

设计并制备了模拟晶界成分偏析效应的厚块合金，并通过冷加工模拟辐照硬化变形。通过裂纹扩展速率（CGR）试验，研究了17Cr-13Ni-3Si、13Cr-17Ni和13Cr-17Ni- 3si合金在铸态和冷加工状态下的应力腐蚀开裂（SCC）行为。三种铸态模拟合金断口均表现出局部SCC特征，而所有冷加工态模拟合金断口均表现出典型的晶间SCC特征。在没有冷加工的情况下，单次添加Si对SCC的扩展没有明显的影响。应力加速晶界氧化是晶间裂纹的起裂部位。冷加工高硅低铬合金在氢化水中显著促进SCC生长。讨论了合金成分和应变硬化对SCC的影响及其对辐照辅助SCC的启示。

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分钟内获得全文去求助

来源期刊

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.