Brittle failure of spent fuel claddings under long-term dry interim storage conditions – Preliminary analysis
- The evaluation of cladding integrity is a major issue to be demonstrated in Germany for extended interim storage periods up to 100 years and subsequent transportation considering operational and accidental conditions with respect to reactor operation, cask drying and dry interim storage. The chemical reaction between the zirconium fuel cladding and the cooling water in water-cooled reactors produces hydrogen and zirconium oxide. Hydrogen diffuses into the cladding and precipitates as zirconium hydrides when the solubility limit is reached, preferably oriented in hoop direction. At high temperatures during vacuum drying procedures, the hydrides can dissolve. Over a succeeding period of slow cooling with existing hoop stress the hydrides precipitate again, but partly reoriented along the radial direction of the cladding. This change of microstructure in combination with a decreasing temperature (0.5...2 K/year) during (extended) interim storage and additional mechanical load by handlingThe evaluation of cladding integrity is a major issue to be demonstrated in Germany for extended interim storage periods up to 100 years and subsequent transportation considering operational and accidental conditions with respect to reactor operation, cask drying and dry interim storage. The chemical reaction between the zirconium fuel cladding and the cooling water in water-cooled reactors produces hydrogen and zirconium oxide. Hydrogen diffuses into the cladding and precipitates as zirconium hydrides when the solubility limit is reached, preferably oriented in hoop direction. At high temperatures during vacuum drying procedures, the hydrides can dissolve. Over a succeeding period of slow cooling with existing hoop stress the hydrides precipitate again, but partly reoriented along the radial direction of the cladding. This change of microstructure in combination with a decreasing temperature (0.5...2 K/year) during (extended) interim storage and additional mechanical load by handling procedures or under accident conditions could lead to a potential cladding embrittlement and consequently increased failure probability. The current research project BRUZL (Fracture mechanical analysis of spent fuel claddings under long-term dry interim storage conditions) has been launched by BAM to investigate potential sudden brittle failure of spent fuel claddings at small deformation under long-term dry interim storage conditions and is presented.…
Autor*innen: | Uwe Zencker, Kai Simbruner, Holger Völzke |
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Dokumenttyp: | Beitrag zu einem Tagungsband |
Veröffentlichungsform: | Graue Literatur |
Sprache: | Englisch |
Titel des übergeordneten Werkes (Englisch): | Proceedings of the IAEA International Conference on the Management of Spent Fuel from Nuclear Power Reactors |
Jahr der Erstveröffentlichung: | 2019 |
Organisationseinheit der BAM: | 3 Gefahrgutumschließungen; Energiespeicher |
3 Gefahrgutumschließungen; Energiespeicher / 3.4 Sicherheit von Lagerbehältern | |
Erste Seite: | Paper IAEA-CN-272/49, 1 |
Letzte Seite: | 8 |
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 Reactors |
Veranstaltungsort: | Vienna, Austria |
Beginndatum der Veranstaltung: | 24.06.2019 |
Enddatum der Veranstaltung: | 28.06.2019 |
Verfügbarkeit des Dokuments: | Datei im Netzwerk der BAM verfügbar ("Closed Access") |
Datum der Freischaltung: | 06.09.2019 |
Referierte Publikation: | Nein |