{"title":"Behavior of FeCrAl-ODS cladding tube under loss-of-coolant accident conditions","authors":"Takafumi Narukawa , Keietsu Kondo , Yuki Fujimura , Kazuo Kakiuchi , Yutaka Udagawa , Yoshiyuki Nemoto","doi":"10.1016/j.jnucmat.2023.154467","DOIUrl":null,"url":null,"abstract":"<div><p><span><span>To evaluate the behavior of an oxide-dispersion-strengthened FeCrAl (FeCrAl-ODS) cladding tube under loss-of-coolant accident (LOCA) conditions of light-water reactors (LWRs), the following two laboratory-scale LOCA-simulated tests were performed: the burst and integral thermal shock tests. Four burst and three integral thermal shock tests were performed on unirradiated, stress-relieved FeCrAl-ODS cladding tube specimens, simulating ballooning and rupture, </span>oxidation, and quenching, which were postulated during a LOCA. The burst temperature of the FeCrAl-ODS cladding tube was 200–</span><figure><img></figure><span><span> higher than that of the Zircaloy<span> cladding tube, and the FeCrAl-ODS cladding tube's maximum circumferential strain was smaller than or equal to that of the Zircaloy-4 cladding tube. These results indicate that the FeCrAl-ODS cladding tube has higher </span></span>strength at high temperatures than the conventional Zircaloy cladding tube. The FeCrAl-ODS cladding tube did not fracture after being subjected to an axial restraint load of ∼</span><figure><img></figure><span>, which is more than 10 times higher than the axial restraint load estimated for existing LWRs, during quenching, following isothermal oxidation at </span><figure><img></figure> for <figure><img></figure>. The FeCrAl-ODS cladding tube was hardly oxidized during this isothermal oxidation condition. However, it melted after a short oxidation at <figure><img></figure> and fractured after abnormal oxidation at <figure><img></figure> for <figure><img></figure>. Based on these results, the FeCrAl-ODS cladding tube should not fracture in the time range expected during LOCAs below <figure><img></figure>, where no melting or abnormal oxidation occurs.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"582 ","pages":"Article 154467"},"PeriodicalIF":2.8000,"publicationDate":"2023-08-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"3","resultStr":null,"platform":"Semanticscholar","paperid":null,"PeriodicalName":"Journal of Nuclear Materials","FirstCategoryId":"5","ListUrlMain":"https://www.sciencedirect.com/science/article/pii/S0022311523002325","RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":null,"EPubDate":"","PubModel":"","JCR":"Q3","JCRName":"MATERIALS SCIENCE, MULTIDISCIPLINARY","Score":null,"Total":0}
引用次数: 3
Abstract
To evaluate the behavior of an oxide-dispersion-strengthened FeCrAl (FeCrAl-ODS) cladding tube under loss-of-coolant accident (LOCA) conditions of light-water reactors (LWRs), the following two laboratory-scale LOCA-simulated tests were performed: the burst and integral thermal shock tests. Four burst and three integral thermal shock tests were performed on unirradiated, stress-relieved FeCrAl-ODS cladding tube specimens, simulating ballooning and rupture, oxidation, and quenching, which were postulated during a LOCA. The burst temperature of the FeCrAl-ODS cladding tube was 200– higher than that of the Zircaloy cladding tube, and the FeCrAl-ODS cladding tube's maximum circumferential strain was smaller than or equal to that of the Zircaloy-4 cladding tube. These results indicate that the FeCrAl-ODS cladding tube has higher strength at high temperatures than the conventional Zircaloy cladding tube. The FeCrAl-ODS cladding tube did not fracture after being subjected to an axial restraint load of ∼, which is more than 10 times higher than the axial restraint load estimated for existing LWRs, during quenching, following isothermal oxidation at for . The FeCrAl-ODS cladding tube was hardly oxidized during this isothermal oxidation condition. However, it melted after a short oxidation at and fractured after abnormal oxidation at for . Based on these results, the FeCrAl-ODS cladding tube should not fracture in the time range expected during LOCAs below , where no melting or abnormal oxidation occurs.
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.
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