{"title":"Effects of magnesium chloride salts on stress corrosion cracking behavior of austenitic stainless steels used in dry storage canister","authors":"Mei-Ya Wang , Ya-Yun Cheng , Tsung-Kuang Yeh","doi":"10.1016/j.nucengdes.2025.114282","DOIUrl":null,"url":null,"abstract":"<div><div>Austenitic stainless steels, including Types 304, 304L, and 316L stainless steel (SS), are commonly adopted canister materials for dry storage of spent nuclear fuels. When spent fuel storage installations are located near chloride-containing areas, stress corrosion cracking of austenitic stainless steels may take place as dried sea salts deposit on the stressed steel surfaces. The purpose of this study is to evaluate the corrosion behaviors of candidate canister materials with U-bend samples exposed to simulated chloride-containing environments. Various test environments were set up in a glass chamber with periodic spraying of magnesium chloride liquid solutions of three different concentrations at various temperatures and a controlled flow of vapor at a constant relative humidity of 40 % for 1500 h. Prior to the exposure tests, all samples underwent treatments of solution annealing and thermal sensitization.</div><div>After each specific test, surface morphologies and the presence of cracks on the samples were examined via scanning electron microscopy analyses. According to the test results in the presence of magnesium chloride deposits, except for the sensitized 304 SS and 304L SS samples, no cracks longer than 500 μm were observed in the sensitized 316L SS sample at 40 °C. The outcome indicated a better corrosion resistance of 316L SS than those of the other two at this designated temperature. At a higher temperature of 60 °C, 304 SS and 304L SS exhibited more cracks and pits than at 40 °C, and coalescence of pits dominated at an even higher temperature of 80 °C. On the other hand, 316L SS showed mainly pitting corrosion at 40 °C, but pits and cracks were observed at 60 °C. In particular, 316L SS exhibited more and deeper localized cracks originating from pits, while a smaller amount of overall corroded surface area was observed at 80 °C.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114282"},"PeriodicalIF":1.9000,"publicationDate":"2025-07-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":null,"platform":"Semanticscholar","paperid":null,"PeriodicalName":"Nuclear Engineering and Design","FirstCategoryId":"5","ListUrlMain":"https://www.sciencedirect.com/science/article/pii/S0029549325004595","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
Austenitic stainless steels, including Types 304, 304L, and 316L stainless steel (SS), are commonly adopted canister materials for dry storage of spent nuclear fuels. When spent fuel storage installations are located near chloride-containing areas, stress corrosion cracking of austenitic stainless steels may take place as dried sea salts deposit on the stressed steel surfaces. The purpose of this study is to evaluate the corrosion behaviors of candidate canister materials with U-bend samples exposed to simulated chloride-containing environments. Various test environments were set up in a glass chamber with periodic spraying of magnesium chloride liquid solutions of three different concentrations at various temperatures and a controlled flow of vapor at a constant relative humidity of 40 % for 1500 h. Prior to the exposure tests, all samples underwent treatments of solution annealing and thermal sensitization.
After each specific test, surface morphologies and the presence of cracks on the samples were examined via scanning electron microscopy analyses. According to the test results in the presence of magnesium chloride deposits, except for the sensitized 304 SS and 304L SS samples, no cracks longer than 500 μm were observed in the sensitized 316L SS sample at 40 °C. The outcome indicated a better corrosion resistance of 316L SS than those of the other two at this designated temperature. At a higher temperature of 60 °C, 304 SS and 304L SS exhibited more cracks and pits than at 40 °C, and coalescence of pits dominated at an even higher temperature of 80 °C. On the other hand, 316L SS showed mainly pitting corrosion at 40 °C, but pits and cracks were observed at 60 °C. In particular, 316L SS exhibited more and deeper localized cracks originating from pits, while a smaller amount of overall corroded surface area was observed at 80 °C.
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.