TY - RPRT A1 - Caruso, Stefano A1 - Vlassopoulos, Efstathios A1 - Dagan, Ron A1 - Fiorito, Luca A1 - Herm, Michel A1 - Jansson, Peter A1 - Kromar, Marjan A1 - Király, Márton A1 - Leppanen, Jaakko A1 - Feria Marquez, Francisco A1 - Metz, Volker A1 - Papaioannou, Dimitrios A1 - Herranz, Luis Enrique A1 - Rochman, Dimitri A1 - Schillebeeckx, Peter A1 - Seidl, Marcus A1 - Hernandez Solis, Augusto A1 - Stankovskiy, Alexey A1 - Alvarez Velarde, Francisco A1 - Verwerft, Marc A1 - Rodriguez Villagra, Nieves A1 - Zencker, Uwe A1 - Žerovnik, Gasper T1 - EURAD - Work Package 8 - Deliverable 8.1 - State-of-the-art report N2 - A state-of-the-art (SOTA) review on characterisation of spent nuclear fuel (SNF) properties in terms of source term and inventory assessment (neutron, gamma-ray emission, decay heat, radionuclide inventory, elemental content) and in terms of out-of-core fuel performance (cladding performance and fuel integrity in view of the safety criteria for SNF interim storage, transport and canister packaging) using several numerical and experimental approaches and methodologies is presented. This SOTA report is a result of the spent fuel characterisation (SFC) work package as part of the European Joint Programme on Radioactive Waste Management (EURAD), which offers an overview of the status of knowledge in the field of SNF characterisation and assessment during the pre-disposal phase. The document aims to focus on the current safety-significant gaps and related challenges, providing a direct link to the goals of the mandated actors of EURAD. The report is expected to be used by all EURAD colleagues in their national programmes as a key resource for knowledge management programmes and to contribute to demonstrating and documenting the state-of-the-art. KW - Radioactive waste management KW - Spent fuel characterisation KW - Extended interim storage KW - Predisposal PY - 2022 UR - https://www.ejp-eurad.eu/publications/eurad-d81-state-art-report SP - 1 EP - 112 PB - Agence Nationale pour la Gestion des Déchets Radioactifs (ANDRA) CY - Châtenay-Malabry AN - OPUS4-59154 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - RPRT A1 - Zencker, Uwe A1 - Aguado, Carlos A1 - Arkoma, Asko A1 - Bertsch, Johannes A1 - Cristóbal Beneyto, Miguel A1 - Feria Marquez, Francisco A1 - Herranz, Luis Enrique A1 - Ivanov, Ivan A1 - Király, Márton A1 - Nasyrow, Ramil A1 - Papaioannou, Dimitrios A1 - Ruiz-Hervias, Jesus A1 - Vlassopoulos, Efstathios T1 - Thermo-mechanical-chemical properties of unirradiated and irradiated samples of spent nuclear fuel rod segments and cladding N2 - The report addresses the experimental characterisation, thermo-mechanical modelling and performance of unirradiated and irradiated samples of spent nuclear fuel rod segments and cladding. Hydrogen is relocated in nuclear fuel claddings by diffusion along temperature, stress, and concentration gradients, and towards a liner if existent. The precipitation of hydrides during cool-down at the beginning of the storage is influenced by mechanical stress. A strong accumulation of hydrides and a disadvantageous hydride alignment in relation to the stress state can affect the mechanical properties and compromise the fuel cladding integrity. Neutron radiography is used to determine spatially resolved hydrogen and hydride distributions in claddings. A liner in duplex claddings attracts hydrogen, which will then no longer be available to form detrimental hydrides in the bulk part of the cladding. Individual three-point bending and gravitational impact tests on filled pressurised spent nuclear fuel rod segments were conducted in hot-cell facilities. The material behaviour under three-point bending in the presence of hydrides and at different temperatures was systematically investigated in quasi-static laboratory tests on unirradiated rod segments with and without surrogate pellets until rupture. Ring compression tests were carried out on irradiated and unirradiated samples at different temperatures and both with and without radial hydrides. With radial hydrides, brittle failure is possible even at low deformation and simultaneously low temperature. Crack initiation will occur on the radial hydride with the highest fracture mechanics load, which depends on the hydride size in the regions with the highest hoop stresses. Due to the high hydrogen content used in creep tests conducted, precipitation hardening of the hydrides in the zirconium alloy significantly reduces the creep rate, as it strengthens the material and inhibits the dislocation sliding. The use of finite element analyses on mechanical tests to determine the cladding properties using a numerical optimisation method based on a large number of automatically generated sensitivity calculations is demonstrated. The brittle fracture of cladding samples with radial hydrides in the ring compression test can be described with a cohesive zone model if the cohesive parameters are selected appropriately and the hydride morphology is specified. Progress was made in the verification, validation and enhancement of spent fuel performance codes by the compilation of a representative validation database (related to rod internal pressure, fission gas release, rod void volume, and in-clad hydrogen distribution), the enhancement of CIEMAT’s FRAPCON-xt code for rod internal pressure prediction at high burnup, the enhancement of CIEMAT’s in-clad hydrogen performance subroutine HYDCLAD with more phenomenological modelling, the extension of INL’s BISON code with a cladding creep law for dry storage, and a benchmark with FRAPCON-xt and BISON simulating in-reactor and dry storage scenarios. KW - Spent Nuclear Fuel KW - Cladding KW - Experimental Characterisation KW - Modelling KW - Fuel Performance Code PY - 2024 SP - 1 EP - 102 PB - European Joint Programme EURAD AN - OPUS4-61008 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - RPRT A1 - Vlassopoulos, Efstathios A1 - Dagan, Ron A1 - Fiorito, Luca A1 - Herm, Michel A1 - Jansson, Peter A1 - Kromar, Marjan A1 - Király, Márton A1 - Leppanen, Jaakko A1 - Feria Marquez, Francisco A1 - Metz, Volker A1 - Papaioannou, Dimitrios A1 - Herranz, Luis Enrique A1 - Rochman, Dimitri A1 - Schillebeeckx, Peter A1 - Seidl, Marcus A1 - Hernandez Solis, Augusto A1 - Stankovskiy, Alexey A1 - Alvarez Velarde, Francisco A1 - Verwerft, Marc A1 - Rodriguez Villagra, Nieves A1 - Wiss, Thierry A1 - Zencker, Uwe A1 - Žerovnik, Gasper T1 - EURAD Work Package 8 Deliverable 8.2 Updated State-of-the-Art Report N2 - The state-of-the-art (SOTA) report offers an overview of the status of knowledge in the area of spent nuclear fuel (SNF) characterisation and assessment during the pre-disposal phase using several numerical and experimental approaches and methodologies. A review on characterisation of SNF properties in terms of source term and inventory assessment (neutron, gamma-ray emission, decay heat, radionuclide inventory, and elemental content) and in terms of out-of-core fuel performance (cladding performance and fuel integrity in view of the safety criteria for SNF interim storage, transport and canister packaging) is presented. This updated SOTA report provides the progress made in the spent fuel characterisation (SFC) work package as part of the European Joint Programme on Radioactive Waste Management (EURAD), documents the identified technical gaps, and provides recommendations for future work. The report is expected to be used by all EURAD colleagues in their national programmes as a key resource for knowledge management programmes and to contribute to demonstrating and documenting the state-of-the-art. KW - Radioactive waste management KW - Spent fuel characterisation KW - Extended interim storage KW - Predisposal PY - 2024 UR - https://www.ejp-eurad.eu/publications/d82-updated-state-art-report SP - 1 EP - 182 AN - OPUS4-61218 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -