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Brittle Failure Limits of Spent Fuel Claddings Subjected to Long-Term Dry Interim Storage Conditions

  • The mechanical properties of spent fuel claddings can be adversely affected under the conditions of long-term dry interim storage, so that the failure limits may be reached in case of mechanical loads during handling or transport after storage. Pre-storage drying and the early stage of interim storage can expose the cladding to higher temperatures and higher tensile hoop stresses than those associated with in-reactor operation and pool storage. During slow cooling of a cladding tube under internal pressure, radial hydrides may precipitate in zirconium-based cladding alloys. This can lead to embrittlement of the material and sudden failure of the cladding integrity under mechanical stress. In order to prevent brittle failure, numerical methods are being developed to predict the mechanical behaviour and identify limiting conditions. Experimental investigations, numerical analyses and evaluation methods are discussed. An established experimental method for characterising claddingThe mechanical properties of spent fuel claddings can be adversely affected under the conditions of long-term dry interim storage, so that the failure limits may be reached in case of mechanical loads during handling or transport after storage. Pre-storage drying and the early stage of interim storage can expose the cladding to higher temperatures and higher tensile hoop stresses than those associated with in-reactor operation and pool storage. During slow cooling of a cladding tube under internal pressure, radial hydrides may precipitate in zirconium-based cladding alloys. This can lead to embrittlement of the material and sudden failure of the cladding integrity under mechanical stress. In order to prevent brittle failure, numerical methods are being developed to predict the mechanical behaviour and identify limiting conditions. Experimental investigations, numerical analyses and evaluation methods are discussed. An established experimental method for characterising cladding materials is the Ring Compression Test (RCT). Some test results on irradiated cladding tubes after operation in pressure water reactors are publicly available. However, it is helpful to carry out studies on unirradiated surrogate claddings with similar material properties to reduce the effort associated with irradiated samples in hot cells and to perform material tests with a wider range of parameters. On the basis of such experimental data, load-displacement curves have been numerically analysed for a selection of cladding materials. Radial hydrides can cause a sample to break suddenly due to fracture even at low deformation. Noticeable load drops in the RCT are caused by unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting adjacent hydrides. The cohesive zone approach was used to simulate the failure process taking into account the radial hydride morphology. The developed method can adequately describe both the deformation and failure behaviour of irradiated as well as unirradiated claddings of zirconium-based alloys with radial hydrides under RCT conditions. Limiting conditions can be expressed in terms of fracture energy and cohesive strength.zeige mehrzeige weniger

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Metadaten
Autor*innen:Uwe Zencker
Koautor*innen:Holger Völzke
Dokumenttyp:Vortrag
Veröffentlichungsform:Präsentation
Sprache:Englisch
Jahr der Erstveröffentlichung:2024
Organisationseinheit der BAM:3 Gefahrgutumschließungen; Energiespeicher
3 Gefahrgutumschließungen; Energiespeicher / 3.4 Sicherheit von Lagerbehältern
DDC-Klassifikation:Technik, Medizin, angewandte Wissenschaften / Ingenieurwissenschaften / Angewandte Physik
Freie Schlagwörter:Ageing Management; Cladding Embrittlement; Extended Storage; Ring Compression Test; Spent Nuclear Fuel
Themenfelder/Aktivitätsfelder der BAM:Energie
Energie / Kerntechnische Entsorgung
Veranstaltung:IAEA International Conference on the Management of Spent Fuel from Nuclear Power Plants - Meeting the Moment
Veranstaltungsort:Vienna, Austria
Beginndatum der Veranstaltung:10.06.2024
Enddatum der Veranstaltung:14.06.2024
Verfügbarkeit des Dokuments:Datei im Netzwerk der BAM verfügbar ("Closed Access")
Datum der Freischaltung:20.06.2024
Referierte Publikation:Nein
Eingeladener Vortrag:Nein
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