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Embrittlement of Spent Fuel Claddings - Micromechanically Motivated Cohesive Zone Modelling

  • Since cladding acts as the first barrier for nuclear fuel, it is key to understand the degrading mechanisms leading to cladding embrittlement. Zirconium alloy based cladding samples subjected to simulated pre-storage drying conditions have shown high susceptibility to brittle failure under the occurrence of hoop stress and low temperature as expected after long-term interim storage. The stress state associated with a compressive pinch-load scenario is reproduced by the Ring Compression Test (RCT), which results in sudden failure at relatively small mechanical loads with high sensitivity to the presence of radial hydrides. As part of the BAM research project BRUZL (Fracture mechanics analysis of spent fuel claddings during long-term dry interim storage), static RCTs were conducted on non-irradiated, pre-hydrided ZIRLO® ring-shaped samples that were subjected to a radial hydride treatment. All samples failed in macroscopically brittle fashion with sudden load drops and severe crackSince cladding acts as the first barrier for nuclear fuel, it is key to understand the degrading mechanisms leading to cladding embrittlement. Zirconium alloy based cladding samples subjected to simulated pre-storage drying conditions have shown high susceptibility to brittle failure under the occurrence of hoop stress and low temperature as expected after long-term interim storage. The stress state associated with a compressive pinch-load scenario is reproduced by the Ring Compression Test (RCT), which results in sudden failure at relatively small mechanical loads with high sensitivity to the presence of radial hydrides. As part of the BAM research project BRUZL (Fracture mechanics analysis of spent fuel claddings during long-term dry interim storage), static RCTs were conducted on non-irradiated, pre-hydrided ZIRLO® ring-shaped samples that were subjected to a radial hydride treatment. All samples failed in macroscopically brittle fashion with sudden load drops and severe crack propagation through almost the entire wall thickness. In a post-RCT examination of the fracture surfaces using scanning electron microscopy, radial hydrides could be identified as the weakest link of the structure. However, the microscopic failure mechanism is much more complex due to cracks initiating at different axial positions in depth direction of the ring, overlapping cracks, and arresting cracks. A considerable portion of the fracture surfaces appeared to have undergone plastic deformation of the zirconium matrix resulting in void growth and coalescence associated with ductile failure. The geometry of the ductile patches was measured, and the ductile surface fraction was determined. Based on these observations, a statistical model was developed that can reproduce the composition of the fracture surface as a one-dimensional projection. A micromechanically motivated cohesive zone model (CZM) was implemented in existing RCT finite element analysis (FEA) models with different constitutive laws for brittle hydrides and ductile matrix respectively. In this work, FEA results are presented for non-irradiated ZIRLO® and high-burnup M5® claddings. The impact of varying hydride-matrix combinations is discussed. Acknowledgement: The project was funded by the former Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561.zeige mehrzeige weniger

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Metadaten
Autor*innen:Kai SimbrunerORCiD
Koautor*innen:Uwe Zencker, Holger Völzke
Dokumenttyp:Vortrag
Veröffentlichungsform:Präsentation
Sprache:Englisch
Jahr der Erstveröffentlichung:2022
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:6th GRS Workshop on the Safety of Extended Dry Storage of Used Nuclear Fuel
Veranstaltungsort:Garching, Germany
Beginndatum der Veranstaltung:01.06.2022
Enddatum der Veranstaltung:03.06.2022
Verfügbarkeit des Dokuments:Datei im Netzwerk der BAM verfügbar ("Closed Access")
Datum der Freischaltung:14.06.2022
Referierte Publikation:Nein
Eingeladener Vortrag:Nein
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