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  <doc>
    <id>58276</id>
    <completedYear/>
    <publishedYear>2023</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst/>
    <pageLast/>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>lecture</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Spent Fuel Characterization - Current Activities at BAM</title>
    <abstract language="eng">The European Joint Programme on Radioactive Waste Management (EURAD) is working on Spent Fuel Characterization (SFC) in its work package (WP) 8. Inspired by the EURAD activities, the International Atomic Energy Agency (IAEA) established an international Coordinated Research Project (CRP) on SFC. The EURAD WP SFC&#13;
participants are collaborating as a team on the IAEA CRP on SFC. The EURAD WP SFC project consists of four tasks. Task 3 investigates the behaviour of nuclear fuel and cladding after discharge. The aim of these activities is to understand and describe the evolution of the cladding-pellet system and its ageing under conditions of extended interim storage, transportation and emplacement in a final disposal system. At a Consultancy Meeting, BAM as contributor to Task 3 presented current results of the failure analysis of irradiated ZIRLO® claddings under conditions of the Ring Compression Test.</abstract>
    <enrichment key="eventName">IAEA Consultancy Meeting on the Coordinated Research Project on Spent Fuel Characterization</enrichment>
    <enrichment key="eventPlace">Online meeting</enrichment>
    <enrichment key="eventStart">12.09.2023</enrichment>
    <enrichment key="eventEnd">13.09.2023</enrichment>
    <enrichment key="opus.source">publish</enrichment>
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    <enrichment key="InvitedTalks">0</enrichment>
    <author>Uwe Zencker</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Nuclear Fuel Cladding</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Numerical Failure Analysis</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ring Compression Test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent Fuel Characterization</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Extended Interim Storage</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>57246</id>
    <completedYear/>
    <publishedYear>2023</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>12</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume>9</volume>
    <type>article</type>
    <publisherName>EDP Sciences</publisherName>
    <publisherPlace>Les Ulis Cedex, France</publisherPlace>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>1</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Spent nuclear fuel management, characterisation, and dissolution behaviour: progress and achievement from SFC and DisCo</title>
    <abstract language="eng">SFC is a work package in Eurad that investigates issues related to the properties of the spent nuclear fuel in the back-end of the nuclear fuel cycle. Decay heat, nuclide inventory, and fuel integrity (mechanical and otherwise), and not least the related uncertainties, are among the primary focal points of SFC. These have very significant importance for the safety and operational aspect of the back-end. One consequence is the operation economy of the back-end, where deeper understanding and quantification allow for significant optimization, meaning that significant parts of the costs can be reduced. In this paper, SFC is described, and examples of results are presented at about half-time of the work package, which will finish in 2024. The DisCo project started in 2017 and finished in November 2021 and was funded under the Horizon 2020 Euratom program. It investigated if the properties of modern fuel types, namely doped fuel, and MOX, cause any significant difference in the dissolution behavior of the fuel matrix compared with standard fuels. Spent nuclear fuel experiments were complemented with studies on model materials as well as the development of models describing the solid state, the dissolution process, and reactive transport in the near field. This research has improved the understanding of processes occurring at the interface between spent nuclear fuel and aqueous solution, such as redox reactions. Overall, the results show that from a long-term fuel matrix dissolution point of view, there is no significant difference between MOX fuel, Cr+Al-doped fuel, and standard fuels.</abstract>
    <parentTitle language="eng">EPJ Nuclear Sciences &amp; Technologies</parentTitle>
    <identifier type="doi">10.1051/epjn/2022029</identifier>
    <identifier type="issn">2491-9292</identifier>
    <identifier type="urn">urn:nbn:de:kobv:b43-572461</identifier>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="date_peer_review">30.03.2023</enrichment>
    <licence>Creative Commons - CC BY - Namensnennung 4.0 International</licence>
    <author>A. Sjöland</author>
    <author>P. Christensen</author>
    <author>L. Zetterström Evins</author>
    <author>D. Bosbach</author>
    <author>L. Duro</author>
    <author>I. Farnan</author>
    <author>V. Metz</author>
    <author>Uwe Zencker</author>
    <author>J. Ruiz-Hervias</author>
    <author>N. Rodriguez-Villagra</author>
    <author>M. Kiraly</author>
    <author>P. Schillebeeckx</author>
    <author>D. Rochman</author>
    <author>M. Seidl</author>
    <author>R. Dagan</author>
    <author>M. Verwerft</author>
    <author>L. E. Herranz Puebla</author>
    <author>D. Hordynskyi</author>
    <author>F. Feria</author>
    <author>E. Vlassopoulos</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent nuclear fuel management</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent fuel characterization</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Dissolution behaviour</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="literaturgattung" number="">Verlagsliteratur</collection>
    <collection role="fulltextaccess" number="">Datei für die Öffentlichkeit verfügbar ("Open Access")</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
    <thesisPublisher>Bundesanstalt für Materialforschung und -prüfung (BAM)</thesisPublisher>
    <file>https://opus4.kobv.de/opus4-bam/files/57246/epjn220033.pdf</file>
  </doc>
  <doc>
    <id>57021</id>
    <completedYear/>
    <publishedYear>2022</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst/>
    <pageLast/>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>lecture</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Container requirements for high-level radioactive waste disposal in rock salt, claystone, and crystalline rock – Outcomes of the research project KoBrA</title>
    <abstract language="eng">Suitable containers for receiving high-level radioactive waste are a key elements of every repository system concerning different host rocks like rock salt, claystone and crystalline rock as being considered in Germany. The disposal container design is decisive regarding boundary conditions for the transport, emplacement techniques, and the operational and long-term safety assessment. The systematic derivation of the requirements to be placed on disposal containers forms the basis for a targeted, comprehensible and transparent development of containers that meet the requirements. The presentation summarizes the outcomes of the research project KoBrA.</abstract>
    <enrichment key="eventName">IGD-TP Symposium</enrichment>
    <enrichment key="eventPlace">Zurich, Switzerland</enrichment>
    <enrichment key="eventStart">20.09.2022</enrichment>
    <enrichment key="eventEnd">22.09.2022</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="InvitedTalks">1</enrichment>
    <author>Holger Völzke</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Disposal</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Container</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Radioactive waste</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Safety</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>61008</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>102</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>report</type>
    <publisherName>European Joint Programme EURAD</publisherName>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Thermo-mechanical-chemical properties of unirradiated and irradiated samples of spent nuclear fuel rod segments and cladding</title>
    <abstract language="eng">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.</abstract>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Uwe Zencker</author>
    <author>Carlos Aguado</author>
    <author>Asko Arkoma</author>
    <author>Johannes Bertsch</author>
    <author>Miguel Cristóbal Beneyto</author>
    <author>Francisco Feria Marquez</author>
    <author>Luis Enrique Herranz</author>
    <author>Ivan Ivanov</author>
    <author>Márton Király</author>
    <author>Ramil Nasyrow</author>
    <author>Dimitrios Papaioannou</author>
    <author>Jesus Ruiz-Hervias</author>
    <author>Efstathios Vlassopoulos</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent Nuclear Fuel</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cladding</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Experimental Characterisation</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Modelling</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Fuel Performance Code</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>61218</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>182</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>report</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">EURAD Work Package 8 Deliverable 8.2 Updated State-of-the-Art Report</title>
    <abstract language="eng">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.</abstract>
    <identifier type="url">https://www.ejp-eurad.eu/publications/d82-updated-state-art-report</identifier>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Efstathios Vlassopoulos</author>
    <author>Ron Dagan</author>
    <author>Luca Fiorito</author>
    <author>Michel Herm</author>
    <author>Peter Jansson</author>
    <author>Marjan Kromar</author>
    <author>Márton Király</author>
    <author>Jaakko Leppanen</author>
    <author>Francisco Feria Marquez</author>
    <author>Volker Metz</author>
    <author>Dimitrios Papaioannou</author>
    <author>Luis Enrique Herranz</author>
    <author>Dimitri Rochman</author>
    <author>Peter Schillebeeckx</author>
    <author>Marcus Seidl</author>
    <author>Augusto Hernandez Solis</author>
    <author>Alexey Stankovskiy</author>
    <author>Francisco Alvarez Velarde</author>
    <author>Marc Verwerft</author>
    <author>Nieves Rodriguez Villagra</author>
    <author>Thierry Wiss</author>
    <author>Uwe Zencker</author>
    <author>Gasper Žerovnik</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Radioactive waste management</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent fuel characterisation</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Extended interim storage</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Predisposal</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>56120</id>
    <completedYear/>
    <publishedYear>2022</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst/>
    <pageLast/>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>lecture</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">EURAD WP 8 Task 3 Overview: Behaviour of nuclear fuel and cladding after discharge</title>
    <abstract language="eng">Task 3 of Work package 8 (Spent Fuel Characterization) of the European Joint Programme on Radioactive Waste Management (EURAD) investigates the behaviour of nuclear fuel and cladding after discharge. The aim of the work is to understand and describe the behaviour of spent nuclear fuel (SNF), irradiated cladding, fuel/cladding chemical interaction (FCCI) and ageing under conditions of extended interim storage, transportation and emplacement in a final disposal system. BAM contributes to the project as partner and leads Task 3. The presentation gives an overview of the project status, main achievements in experimental work and modelling studies, deviations from the plan, delays and challenges ahead.</abstract>
    <enrichment key="eventName">EURAD Work Package 8 (SFC) Annual Meeting</enrichment>
    <enrichment key="eventPlace">Kalmar, Sweden</enrichment>
    <enrichment key="eventStart">19.09.2022</enrichment>
    <enrichment key="eventEnd">21.09.2022</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <enrichment key="InvitedTalks">0</enrichment>
    <author>Uwe Zencker</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Nuclear fuel</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cladding</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Discharge</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>56121</id>
    <completedYear/>
    <publishedYear>2022</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst/>
    <pageLast/>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>lecture</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">EURAD WP 8 (SFC) Task 3 - Update: Behaviour of nuclear fuel and cladding after discharge</title>
    <abstract language="eng">The European Joint Programme on Radioactive Waste Management (EURAD) is working on Spent Fuel Characterization (SFC) in its work package (WP) 8. Inspired by the EURAD activities, the International Atomic Energy Agency (IAEA) established an international Coordinated Research Project (CRP) on SFC. The EURAD WP SFC participants are collaborating as a team on the IAEA CRP on SFC. The EURAD WP SFC project consists of four tasks. Task 3 investigates the behaviour of nuclear fuel and cladding after discharge. The aim of these activities is to understand and describe the evolution of the cladding-pellet system and its ageing under conditions of extended interim storage, transportation and emplacement in a final disposal system. At the Second Research Coordination Meeting on Spent Fuel Characterization, BAM as leader of Task 3 reported on the status of the research work.</abstract>
    <enrichment key="eventName">IAEA 2nd Research Coordination Meeting of the Coordinated Research Project on Spent Fuel Characterization</enrichment>
    <enrichment key="eventPlace">Kalmar, Sweden</enrichment>
    <enrichment key="eventStart">20.09.2022</enrichment>
    <enrichment key="eventEnd">23.09.2022</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <enrichment key="InvitedTalks">0</enrichment>
    <author>Uwe Zencker</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Nuclear fuel</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cladding</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Discharge</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>56853</id>
    <completedYear/>
    <publishedYear>2022</publishedYear>
    <thesisYearAccepted/>
    <language>deu</language>
    <pageFirst>1</pageFirst>
    <pageLast>106</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>report</type>
    <publisherName>Bundesanstalt für Materialforschung und -prüfung (BAM)</publisherName>
    <publisherPlace>Berlin</publisherPlace>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="deu">Entwicklung eines bruchmechanischen Berechnungsansatzes zur Beschreibung des Festigkeitsverhaltens von Brennstabhüllrohren bei längerfristiger Zwischenlagerung (BRUZL)</title>
    <abstract language="deu">Es werden Methoden zur sicherheitstechnischen Bewertung des mechanischen Verhaltens von Brennstabhüllrohren entwickelt, um sprödes Versagen in Belastungsszenarien nach längerfristiger trockener Zwischenlagerung zu verhindern. Unbestrahlte Brennstabhüllrohrsegmente aus der Zirkoniumlegierung ZIRLO® wurden hydriert und thermomechanischen Behandlungen unterzogen, um Hüllrohrmaterial durch Ausscheidung radialer Hydride zu verspröden. Die Proben mit radialen Hydriden wurden im Ring Compression Test (RCT) bei Raumtemperatur untersucht. Bereits bei kleinen Probenverformungen wurden abrupte Kraftabfälle mit instabiler Rissausbreitung durch das Netzwerk radialer Hydride beobachtet. Die Risslängen erreichten in einigen Fällen bis zu 90 % der Wanddicke. Der mikromechanische Versagensmechanismus ist Quasi-Spaltbruch in den Hydriden und Porenbildung, -wachstum und -vereinigung in der Zirkoniummatrix an den duktilen Brücken zwischen benachbarten Hydriden. Es wurde ein Finite-Elemente-Modell des Ring Compression Tests erstellt, um das Versagensverhalten zirkoniumbasierter Legierungen mit radialen Hydriden im RCT zu analysieren. Die Fließkurve für das elastisch-plastische Materialmodell des Matrixmaterials wurde durch eine inverse Finite-Elemente-Analyse ermittelt. Kohäsivzonenmodelle wurden benutzt, um die plötzlichen Kraftabfälle zu reproduzieren. Es wurde ein mikromechanisches Modell mit statistisch generierten Matrix-Hydrid-Verteilungen entwickelt, das zwischen sprödem Versagen entlang der Hydride und duktilem Versagen der Zirkoniummatrix in der Kohäsivzone unterscheidet. Für das jeweilige Materialverhalten wird ein lokales zweiparametriges Versagenskriterium auf Grundlage der Kohäsionsfestigkeit und der Separationsenergie bei einer festgelegten Form des Kohäsivgesetzes vorgeschlagen. Geeignete Kohäsivparameter lassen sich mit einer inversen Finite-Elemente-Analyse des Versagensverhaltens von hydrierten Proben im Ring Compression Test bestimmen. In den Simulationen lag der Schwerpunkt auf dem ersten Kraftabfall. Die Berechnungsergebnisse stimmen gut mit den RCT-Ergebnissen überein. Es konnte gezeigt werden, dass die Rissinitiierung und -ausbreitung stark von der speziellen Anordnung der radialen Hydride und Matrixbrücken in der Bruchzone abhängt. Die numerische Modellbildung wurde an Versuchen mit bestrahlten Proben aus der Zirkoniumlegierung M5® validiert.</abstract>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Uwe Zencker</author>
    <author>Kai Simbruner</author>
    <author>Holger Völzke</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Zwischenlagerung</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Brennstabhüllrohr</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Versagensmechanismus</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Kohäsivzonenmodell</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Ring Compression Test</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>60120</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst/>
    <pageLast/>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>poster</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Consequences of an Imperfectly Mounted Reinforcement Cage in a Cylindrical Concrete Container During Mechanical Specimen Tests</title>
    <abstract language="eng">In 2007 the license for the German Konrad repository for low and intermediate level radioactive waste was finally confirmed. Since then, the site is improved, prepared and equipped for the currently planned start of waste package emplacement in 2029. Bundesanstalt fuer Materialforschung und -pruefung is regularly contracted by the Bundesgesellschaft für Endlagerung as the responsible authority for waste product control, waste container design testing, and waste package acceptance. In this context BAM is mainly responsible for design testing of various container types including the evaluation of quality assurance measures for container manufacturing.&#13;
In addition to general requirements concerning container design, the casks have to withstand specific mechanical load scenarios. Drop tests from different heights under consideration of the most damaging drop orientation and a stacking test are two examples for necessary safety demonstrations.&#13;
If the containers are made from concrete the reinforcement has the function to absorb tensile forces the concrete is not able to withstand by its own. Because of manufacturing imperfections of cylindrical concrete containers manufactured already decades ago, the reinforcement cages are not always exactly, asymmetrically positioned when they were poured. This inspired BAM to perform various ABAQUS® finite-element (FE) simulations by using the example of a simplified cylindrical container design with generic dimensions. On the basis of the calculated tensile and compressive stress distributions suggestions are derived on how a cylindrical concrete container with an imperfectly placed reinforcement cage has to be oriented for a drop or stacking test to ensure the most damaging test scenario.</abstract>
    <enrichment key="eventName">RAMTRANS 2024</enrichment>
    <enrichment key="eventPlace">London,  United Kingdom</enrichment>
    <enrichment key="eventStart">14.05.2024</enrichment>
    <enrichment key="eventEnd">17.05.2024</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Mike Weber</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Finite-element-analysis</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Container design testing procedure</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Reinforced concrete containers</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Drop test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Stacking test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Imperfectly mounted reinforcement cage</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>60323</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>10</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Brittle Failure Limits of Spent Fuel Claddings Subjected to Long-Term Dry Interim Storage Conditions</title>
    <abstract language="eng">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.&#13;
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.&#13;
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.&#13;
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.</abstract>
    <parentTitle language="eng">Proceedings of the IAEA International Conference on the Management of Spent Fuel from Nuclear Power Plants - Meeting the Moment</parentTitle>
    <enrichment key="eventName">IAEA International Conference on the Management of Spent Fuel from Nuclear Power Plants - Meeting the Moment</enrichment>
    <enrichment key="eventPlace">Vienna, Austria</enrichment>
    <enrichment key="eventStart">10.06.2024</enrichment>
    <enrichment key="eventEnd">14.06.2024</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Uwe Zencker</author>
    <author>Holger Völzke</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ageing Management</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cladding Embrittlement</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Extended Storage</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ring Compression Test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent Nuclear Fuel</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>60324</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst/>
    <pageLast/>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>lecture</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Brittle Failure Limits of Spent Fuel Claddings Subjected to Long-Term Dry Interim Storage Conditions</title>
    <abstract language="eng">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.&#13;
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.&#13;
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.&#13;
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.</abstract>
    <enrichment key="eventName">IAEA International Conference on the Management of Spent Fuel from Nuclear Power Plants - Meeting the Moment</enrichment>
    <enrichment key="eventPlace">Vienna, Austria</enrichment>
    <enrichment key="eventStart">10.06.2024</enrichment>
    <enrichment key="eventEnd">14.06.2024</enrichment>
    <enrichment key="InvitedTalks">0</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Uwe Zencker</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ageing Management</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cladding Embrittlement</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Extended Storage</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ring Compression Test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent Nuclear Fuel</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>60668</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst/>
    <pageLast/>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>lecture</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Fracture Mechanics Analysis of Spent Fuel Claddings during Long-Term Dry Interim Storage</title>
    <abstract language="eng">The European Joint Programme on Radioactive Waste Management (EURAD) is working on Spent Fuel Characterization (SFC) in its work package (WP) 8. Inspired by the EURAD activities, the International Atomic Energy Agency (IAEA) established an international Coordinated Research Project (CRP) on SFC. The EURAD WP SFC participants are collaborating as a team on the IAEA CRP on SFC. The EURAD WP SFC project consists of four tasks. Task 3 investigates the behaviour of nuclear fuel and cladding after discharge. The aim of these activities is to understand and describe the evolution of the cladding-pellet system and its ageing under conditions of extended interim storage, transportation and emplacement in a final disposal system. At the Third Research Coordination Meeting on Spent Fuel Characterization, BAM as contributor to Task 3 reported on the activities and status of its research work.</abstract>
    <enrichment key="eventName">IAEA 3rd Research Coordination Meeting of the Coordinated Research Project on Spent Fuel Characterization</enrichment>
    <enrichment key="eventPlace">Cockermouth, UK</enrichment>
    <enrichment key="eventStart">01.07.2024</enrichment>
    <enrichment key="eventEnd">05.07.2024</enrichment>
    <enrichment key="InvitedTalks">0</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Uwe Zencker</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Nuclear fuel</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cladding</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Discharge</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>60671</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>8</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Failure Analysis on Irradiated Claddings Subjected to Long-Term Dry Interim Storage</title>
    <abstract language="eng">Long-term dry interim storage may adversely affect the mechanical properties of spent fuel rods, possibly resulting in a reduced resilience during handling or transport after storage. The cladding is the first barrier for the spent fuel pellets. An established method for characterising the cladding material is the ring compression test (RCT), in which a small, cylindrical sample of the cladding tube is subjected to a compressive load. Radial hydrides may precipitate in zirconium-based alloys (Zircaloy) under pre-storage drying and during slow cooling, which result in embrittlement of the cladding material and eventually a possible sudden failure of cladding integrity under additional mechanical loads. The focus of the presented research is on the development of appropriate nu-merical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. A modelling approach based on cohesive zones is ex-plained which is able to reproduce the propagation of cracks initiated at radial hydrides in the zir-conium matrix. The developed methods are applied to defueled samples of cladding alloy ZIRLO®, which were subjected to a thermo-mechanical treatment to reorient existing circumferential hy-drides to radial hydrides. A selected sample showing sudden load drops during a quasi-static ring compression test is analysed by means of fracture mechanics for illustrative purposes. Based on the developed fracture mechanics approach, not only the deformation behaviour but also the fail-ure behaviour of irradiated as well as unirradiated Zircaloy claddings with radial hydrides under RCT loading conditions can be adequately described.</abstract>
    <parentTitle language="eng">SMiRT 27 Conference Proceedings</parentTitle>
    <enrichment key="eventName">27th International Conference on Structural Mechanics in Reactor Technology - SMiRT 27</enrichment>
    <enrichment key="eventPlace">Yokohama, Japan</enrichment>
    <enrichment key="eventStart">03.03.2024</enrichment>
    <enrichment key="eventEnd">08.03.2024</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Uwe Zencker</author>
    <author>Mohan Reddy Gaddampally</author>
    <author>Holger Völzke</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cladding Embrittlement</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cohesive Zone Modelling</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ring Compression Test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent Nuclear Fuel</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>60672</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst/>
    <pageLast/>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>lecture</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Failure Analysis on Irradiated Claddings Subjected to Long-Term Dry Interim Storage</title>
    <abstract language="eng">Long-term dry interim storage may adversely affect the mechanical properties of spent fuel rods, possibly resulting in a reduced resilience during handling or transport after storage. The cladding is the first barrier for the spent fuel pellets. An established method for characterising the cladding material is the ring compression test (RCT), in which a small, cylindrical sample of the cladding tube is subjected to a compressive load. Radial hydrides may precipitate in zirconium-based alloys (Zircaloy) under pre-storage drying and during slow cooling, which result in embrittlement of the cladding material and eventually a possible sudden failure of cladding integrity under additional mechanical loads. The focus of the presented research is on the development of appropriate nu-merical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. A modelling approach based on cohesive zones is ex-plained which is able to reproduce the propagation of cracks initiated at radial hydrides in the zir-conium matrix. The developed methods are applied to defueled samples of cladding alloy ZIRLO®, which were subjected to a thermo-mechanical treatment to reorient existing circumferential hy-drides to radial hydrides. A selected sample showing sudden load drops during a quasi-static ring compression test is analysed by means of fracture mechanics for illustrative purposes. Based on the developed fracture mechanics approach, not only the deformation behaviour but also the fail-ure behaviour of irradiated as well as unirradiated Zircaloy claddings with radial hydrides under RCT loading conditions can be adequately described.</abstract>
    <enrichment key="eventName">27th International Conference on Structural Mechanics in Reactor Technology - SMiRT 27</enrichment>
    <enrichment key="eventPlace">Yokohama, Japan</enrichment>
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    <author>Uwe Zencker</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cladding Embrittlement</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cohesive Zone Modelling</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ring Compression Test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent Nuclear Fuel</value>
    </subject>
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    <title language="eng">Consequences of an imperfectly mounted reinforcement cage in a generical cylindrical concrete container during mechanical specimen tests</title>
    <abstract language="eng">By highest court decision in 2007 the issued license for the Konrad repository (a former iron ore mine) for low and intermediate level radioactive waste was finally confirmed. Since then, the site is improved, prepared and equipped for the currently planned start of waste package emplacement by 2029. The Bundesanstalt fuer Materialforschung und -pruefung (BAM, Federal Institute for Materials Research and Testing) is regularly contracted by the Bundesgesellschaft für Endlagerung (BGE, Federal Company for Radioactive Waste Disposal) as the responsible authority for waste product control, waste container design testing, and waste package acceptance. In this context BAM is mainly responsible for design testing of various container types including the evaluation quality assurance measures for container manufacturing.&#13;
In the written Konrad disposal requirements [1] [2], all acceptance criteria for potential final disposal containers are defined. In addition to general requirements concerning container design, materials, dimensions and others, specific mechanical load scenarios the casks have to withstand are derived from an analysis of the on-site handling and emplacement procedures. Drop tests from different heights under consideration of the most damaging drop orientation and a stacking test are two examples for safety demonstrations that have to be performed during the container design testing procedure to qualify the respective container type.&#13;
The containers, whether cubic or cylindrical containers, can be made from different materials like ductile cast iron, reinforced concrete or as welded construction from steel plates and profiles. In case of concrete containers the reinforcement has the function to absorb tensile forces the concrete is not able to withstand by its own. Because of manufacturing imperfections of cylindrical concrete waste containers manufactured already decades ago, the reinforcement cages are not always exactly, axisymmetrically positioned when they were poured. This inspired BAM to perform various ABAQUS finite-element (FE) simulations [3] by using the example of a simplified cylindrical container design with generic dimensions and pure elastic material properties. As outcome the stress conditions in the container during the drop test impact under different drop positions and during the stacking test were investigated and are explained and illustrated in this contribution. Different reinforcement cage positions were calculated and compared. On the basis of the calculated tensile and compressive stress distributions suggestions are derived on how a cylindrical concrete container with an imperfectly placed reinforcement cage has to be oriented for a drop or stacking test to ensure the most damaging test scenario.</abstract>
    <enrichment key="eventName">ASME 2024 Pressure Vessels &amp; Piping Conference</enrichment>
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    <author>Mike Weber</author>
    <subject>
      <language>eng</language>
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      <value>Container design testing procedure</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Reinforced concrete containers</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Drop test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Stacking test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Imperfectly mounted reinforcement cage</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Finite-element-analysis</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
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    <publishedYear>2024</publishedYear>
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    <publisherName>Walter de Gruyter GmbH</publisherName>
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    <title language="eng">Brittle failure analysis and modeling of high-burnup PWR fuel cladding alloys</title>
    <abstract language="eng">The aim of this research is the development of methods for predicting mechanical behavior and identification of limiting conditions to prevent brittle failure of high-burnup (HBU) pressure water reactor (PWR) fuel cladding alloys. A finite element (FE) model of the ring compression test (RCT) was created to analyze the failure behavior of zirconium-based alloys with radial hydrides during the RCT. An elastic-plastic material model describes the zirconium alloy. The stress-strain curve needed for the elastic-plastic material model was derived by inverse finite element analyses. Cohesive zone modeling is used to reproduce sudden load drops during RCT loading. Based on the failure mechanism in non-irradiated ZIRLO® claddings, a micro-mechanical model was developed that distinguishes between brittle failure along hydrides and ductile failure of the zirconium matrix. Two different cohesive laws representing these types of failure are present in the same cohesive interface. The key differences between these constitutive laws are the cohesive strength, the stress at which damage initiates, and the cohesive energy, which is the damage energy dissipated by the cohesive zone. Statistically generated matrix-hydride distributions were mapped onto the cohesive elements and simulations with focus on the first load drop were performed. Computational results are in good agreement with the RCT results conducted on high-burnup M5® samples. It could be shown that crack initiation and propagation strongly depend on the specific configuration of hydrides and matrix material in the fracture area.</abstract>
    <parentTitle language="eng">Kerntechnik</parentTitle>
    <identifier type="doi">10.1515/kern-2024-0109</identifier>
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A finite element (FE) model of the ring compression test (RCT) was created to analyze the failure behavior of zirconium-based alloys with radial hydrides during the RCT. An elastic-plastic material model describes the zirconium alloy. The stress-strain curve needed for the elastic-plastic material model was derived by inverse finite element analyses. Cohesive zone modeling is used to reproduce sudden load drops during RCT loading. Based on the failure mechanism in non-irradiated ZIRLO&lt;jats:sup&gt;\u00ae&lt;\/jats:sup&gt; claddings, a micro-mechanical model was developed that distinguishes between brittle failure along hydrides and ductile failure of the zirconium matrix. Two different cohesive laws representing these types of failure are present in the same cohesive interface. The key differences between these constitutive laws are the cohesive strength, the stress at which damage initiates, and the cohesive energy, which is the damage energy dissipated by the cohesive zone. Statistically generated matrix-hydride distributions were mapped onto the cohesive elements and simulations with focus on the first load drop were performed. Computational results are in good agreement with the RCT results conducted on high-burnup M5&lt;jats:sup&gt;\u00ae&lt;\/jats:sup&gt; samples. 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    <author>Kai Simbruner</author>
    <author>Michael C. Billone</author>
    <author>Uwe Zencker</author>
    <author>Yung Y. Liu</author>
    <author>Holger Völzke</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cladding</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Radial hydrides</value>
    </subject>
    <subject>
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      <type>uncontrolled</type>
      <value>Ring compression test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cohesive zone model</value>
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    <title language="eng">Deep geological disposal - Editorial to the topical collection</title>
    <abstract language="eng">The subsurface geological environment, historically a key source of fossil and mineral resources, has seen a significant broadening of its potential applications in recent years, with increasing emphasis on its use for energy production and storage.&#13;
A critical factor in utilizing subsurface space is ensuring the sustainability of management practices, with particular attention to the potential environmental impacts that may result.&#13;
The current Topical Collection on Deep Geological Disposal places greater emphasis on the geotechnical aspects of geological disposal, with a particular focus on the safe isolation of hazardous materials, including radioactive waste. The Topical Collection on Deep Geological Disposal covers both experimental (laboratory and in-situ) and theoretical studies of various geotechnical applications, such as geological disposal of radioactive or other materials, e.g. coal-based solid waste, carbon dioxide sequestration in fractured reservoirs, and brine solutions, but also aspects of uranium mining.&#13;
The articles in the Topical Collections are arranged under &#13;
the following headings:&#13;
– Fundamental studies&#13;
– Experimental studies&#13;
– Numerical studies</abstract>
    <parentTitle language="eng">Environmental Earth Sciences</parentTitle>
    <identifier type="doi">10.1007/s12665-024-12075-2</identifier>
    <identifier type="issn">1866-6299</identifier>
    <identifier type="urn">urn:nbn:de:kobv:b43-624955</identifier>
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    <licence>Creative Commons - CC BY - Namensnennung 4.0 International</licence>
    <author>Thomas Nagel</author>
    <author>Wolfram Rühaak</author>
    <author>Florian Amann</author>
    <author>Guido Bracke</author>
    <author>Stefan Buske</author>
    <author>Julia Kowalski</author>
    <author>Sönke Reiche</author>
    <author>Thorsten Schäfer</author>
    <author>Traugott Scheytt</author>
    <author>Thorsten Stumpf</author>
    <author>Holger Völzke</author>
    <author>Florian Wellmann</author>
    <author>Olaf Kolditz</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Disposal</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Radioactive Waste</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Geology</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
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    <title language="eng">Consequences of an imperfectly mounted reinforcement cage in a generical cylindrical concrete container during mechanical specimen tests</title>
    <abstract language="eng">By highest court decision in 2007 the issued license for the Konrad repository (a former iron ore mine) for low and intermediate level radioactive waste was finally confirmed. Since then, the site is improved, prepared and equipped for the currently planned start of waste package emplacement by 2029. The Bundesanstalt fuer Materialforschung und -pruefung (BAM, Federal Institute for Materials Research and Testing) is regularly contracted by the Bundesgesellschaft für Endlagerung (BGE, Federal Company for Radioactive Waste Disposal) as the responsible authority for waste product control, waste container design testing, and waste package acceptance. In this context BAM is mainly responsible for design testing of various container types including the evaluation quality assurance measures for container manufacturing.&#13;
In the written Konrad disposal requirements all acceptance criteria for potential final disposal containers are defined. In addition to general requirements concerning container design, materials, dimensions and others, specific mechanical load scenarios the casks have to withstand are derived from an analysis of the on-site handling and emplacement procedures. Drop tests from different heights under consideration of the most damaging drop orientation and a stacking test are two examples for safety demonstrations that have to be performed during the container design testing procedure to qualify the respective container type.&#13;
The containers, whether cubic or cylindrical containers, can be made from different materials like ductile cast iron, reinforced concrete or as welded construction from steel plates and profiles. In case of concrete containers the reinforcement has the function to absorb tensile forces the concrete is not able to withstand by its own. Because of manufacturing imperfections of cylindrical concrete waste containers manufactured already decades ago, the reinforcement cages are not always exactly, axisymmetrically positioned when they were poured. This inspired BAM to perform various ABAQUS finite-element (FE) simulations by using the example of a simplified cylindrical container design with generic dimensions and pure elastic material properties. As outcome the stress conditions in the container during the drop test impact under different drop positions and during the stacking test were investigated and are explained and illustrated in this contribution. Different reinforcement cage positions were calculated and compared. On the basis of the calculated tensile and compressive stress distributions suggestions are derived on how a cylindrical concrete container with an imperfectly placed reinforcement cage has to be oriented for a drop or stacking test to ensure the most damaging test scenario.</abstract>
    <parentTitle language="eng">Conference Proceedings ASME 2024 Pressure Vessels &amp; Piping Conference</parentTitle>
    <identifier type="isbn">978-0-7918-8851-3</identifier>
    <identifier type="doi">10.1115/PVP2024-123797</identifier>
    <enrichment key="eventName">ASME 2024 Pressure Vessels &amp; Piping Conference</enrichment>
    <enrichment key="eventPlace">Bellevue, Washington, USA</enrichment>
    <enrichment key="eventStart">28.07.2024</enrichment>
    <enrichment key="eventEnd">02.08.2024</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="date_peer_review">26.05.2025</enrichment>
    <author>Mike Weber</author>
    <author>Holger Völzke</author>
    <author>Gregor Nieslony</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Container design testing procedure</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Reinforced concrete containers</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Drop test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Stacking test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Imperfectly mounted reinforcement cage</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Finite-element-analysis</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>61581</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>deu</language>
    <pageFirst>1</pageFirst>
    <pageLast>188</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>doctoralthesis</type>
    <publisherName>DepositOnce</publisherName>
    <publisherPlace>Berlin</publisherPlace>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>2024-07-03</thesisDateAccepted>
    <title language="deu">Beitrag zur numerischen Beanspruchungsanalyse an dünnwandigen Stahlblechcontainern im Fallversuch - Schädigungsmechanische Materialmodellierung duktiler Werkstoffe unter Berücksichtigung der Dehnratenabhängigkeit der plastischen Instabilität</title>
    <abstract language="deu">Gegenstand der vorliegenden Arbeit ist die Bereitstellung expliziter numerischer Simulationen von Fallversuchen mit dünnwandigen Stahlblechcontainern, die als Endlagerbehälter für radioaktive Abfälle mit vernachlässigbarer Wärmeentwicklung, d. h. schwach- und mittelradioaktive Abfälle, für die Verbringung in das Endlager Konrad vorgesehen sind. Die Betrachtung von potentiellen Absturzszenarien ist ein grundlegender und zentraler Bestandteil der Sicherheitsnachweise zur Auslegung der verschiedenen endlagergängigen Behälter- und Containertypen gegen mechanische Störfallbelastungen. Für dünnwandige Stahlblechcontainer existierten bislang keine sicherheitstechnischen Untersuchungen, die auf numerischen Bewertungsmethoden basieren und eine Bewertung der vorhandenen Sicherheitsreserven hinsichtlich der mechanischen Integrität ermöglichen. Daraus resultierte die Motivation, ein verifiziertes Finite-Elemente-Modell eines repräsentativen Referenzcontainers zu entwickeln, das das dynamische Verformungs- sowie duktile Schädigungs- und Versagensverhalten in numerischen Simulationen von dynamischen Lastfällen physikalisch mit der erforderlichen Güte beschreibt.&#13;
&#13;
Duktile Schädigungs- und Versagensvorgänge metallischer Werkstoffe gehen grundsätzlich mit großen plastischen Dehnungen einher und führen somit unweigerlich zu der Notwendigkeit der Diskussion von plastischer Instabilität und damit verknüpften Lokalisierungseffekten, die einen wesentlichen Einfluss auf die Entwicklung des lokalen Beanspruchungszustands haben. Für die schädigungsmechanische Materialmodellierung folgt hieraus das grundsätzliche Erfordernis der korrekten Berücksichtigung des Instabilitätsbeginns, der in der Regel dehnratenunabhängig angenommen wird. Für dynamische Randbedingungen ist der werkstoffspezifische Einfluss der Dehnrate auf das Einsetzen der Lokalisierung jedoch nicht uneingeschränkt vernachlässigbar, so dass im Rahmen dieser Arbeit ein theoretischer Ansatz ausgearbeitet wird, der die plastische Instabilität dehnratenabhängiger Werkstoffe anhand des zeit- und ortsabhängigen Dehnungsgradienten im einachsigen Zugversuch diskutiert. Darüber hinaus ist für eine hinreichend genaue Materialmodellierung neben der korrekten Prognose des Instabilitätsbeginns auch die korrekte Beschreibung des Verformungsverhaltens jenseits der Gleichmaßdehnung notwendig. Zu diesem Zweck wird in dieser Arbeit eine konstitutive Beziehung entwickelt, die sowohl unter Berücksichtigung der plastischen Instabilität als auch unter der Beachtung von ver- sowie entfestigenden Einflussfaktoren der dynamischen Plastizität formuliert werden kann.&#13;
&#13;
Insgesamt wird eine praxistaugliche Bewertungsmethode der mechanischen Integrität dünnwandiger Stahlblechcontainer unter Störfallbedingungen auf Basis expliziter dynamischer Simulationen erarbeitet, die die komplexen duktilen Schädigungs- und Versagensmechanismen dehnratenabhängiger Werkstoffe berücksichtigt. Mit der zur Simulation von Fallversuchen entwickelten numerischen Modellbildung und den zur Verifizierung entsprechend mit Testcontainern im Originalmaßstab durchgeführten Fallversuchen wird nachgewiesen, dass anhand des Finite-Elemente-Modells eines repräsentativen Referenzcontainers zum einen eine quantitativ belegte Bestimmung ungünstiger Fallorientierungen hinsichtlich maximaler Verformungen und zum anderen eine realistische Abbildung des duktilen Versagensverhaltens ermöglicht wird. Damit können tatsächlich vorhandene Sicherheitsreserven bei potentiell zu unterstellenden Behälterabstürzen quantitativ aufgezeigt und zukünftig eine verbesserte sicherheitstechnische Beurteilung von Stahlblechcontainern im Fallversuch bereitgestellt werden. Damit leistet die vorliegende Arbeit einen wichtigen Beitrag für die Bewertung von Sicherheitsnachweisen von Endlagerbehältern für radioaktive Abfälle und trägt dazu bei, den langfristigen Schutz von Mensch und Umwelt vor Schäden durch radioaktive Stoffe zu gewährleisten.</abstract>
    <abstract language="eng">The subject of the present work is the provision of explicit numerical simulations of drop tests with thin-walled sheet steel containers, which are intended as repository containers for radioactive waste with negligible heat generation, i.e. low and intermediate level waste, for shipment to the Konrad repository. The consideration of potential crash scenarios is a fundamental and central component of the safety analyses for the design of the various types of casks and containers suitable for final disposal against mechanical accident loads. For thin-walled sheet steel containers, no safety investigations based on numerical assessment methods existed so far that would allow an evaluation of the existing safety reserves with regard to mechanical integrity. This resulted in the motivation to develop a verified finite element model of a representative reference container that physically describes the dynamic deformation and ductile damage and failure behaviour in numerical simulations of dynamic load cases with the required quality.&#13;
&#13;
Ductile damage and failure processes of metallic materials are fundamentally accompanied by large plastic strains and thus inevitably lead to the need to discuss plastic instability and associated localisation effects, which have a significant influence on the development of the local stress state. For damage-mechanical material modelling this results in the fundamental requirement of correctly considering the onset of instability, which is usually assumed to be strain-rate independent. For dynamic boundary conditions, however, the material-specific influence of the strain rate on the onset of localisation cannot be neglected without restriction, so that a theoretical approach is elaborated within the scope of this work that discusses the plastic instability of strain-rate-dependent materials on the basis of the time- and location-dependent strain gradient in the uniaxial tensile test. Furthermore, for a sufficiently accurate material modelling, besides the correct prediction of the onset of instability, the correct description of the deformation behaviour in the postcritical region, i.e., beyond the uniform strain, is necessary. For this purpose, a constitutive relationship is developed in this work, which can be formulated both under consideration of the plastic instability as well as under consideration of hardening as well as softening influencing factors of the dynamic plasticity.&#13;
&#13;
Overall, a practical assessment method of the mechanical integrity of thin-walled sheet steel containers under accident conditions is developed based on explicit dynamic simulations, which takes into account the complex ductile damage and failure mechanisms of strain rate-dependent materials. With the numerical modelling developed for the simulation of drop tests and the drop tests carried out accordingly with full-scale test containers for verification, it is demonstrated that, on the basis of the finite element model of a representative reference container, on the one hand a quantitatively proven determination of unfavourable drop orientations with regard to maximum deformations and, on the other hand, a realistic representation of the ductile failure behaviour is made possible. This allows it to quantitatively demonstrate actually existing safety reserves in case of potential container crashes and to provide an improved safety assessment of sheet steel containers in drop tests in the future. This work thus makes an important contribution to the evaluation of safety verifications of final storage containers for radioactive waste and helps to ensure the long-term protection of humans and the environment from damage caused by radioactive substances.</abstract>
    <identifier type="doi">10.14279/depositonce-21874</identifier>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <licence>Creative Commons - CC BY - Namensnennung 4.0 International</licence>
    <author>Christian Keller</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>FEM</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Dynamische Plastizität</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Plastische Instabilität</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Duktile Schädigung</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Endlager Konrad</value>
    </subject>
    <collection role="ddc" number="620">Ingenieurwissenschaften und zugeordnete Tätigkeiten</collection>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="themenfelder" number="">Material</collection>
    <collection role="themenfelder" number="">Degradationsmechanismen</collection>
    <collection role="fulltextaccess" number="">Datei für die Öffentlichkeit verfügbar ("Open Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
    <collection role="unnumberedseries" number="">BAM Dissertationen ohne Nummerierung</collection>
    <thesisPublisher>Bundesanstalt für Materialforschung und -prüfung (BAM)</thesisPublisher>
    <thesisGrantor>Technische Universität Berlin</thesisGrantor>
  </doc>
  <doc>
    <id>57705</id>
    <completedYear/>
    <publishedYear>2023</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst/>
    <pageLast/>
    <pageNumber/>
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    <volume/>
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    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
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    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">System of Quality Assurance Measures in Manufacturing and Operation of Dual Purpose Casks for Spent Nuclear Fuel and High Level Waste in Germany</title>
    <abstract language="eng">German spent nuclear fuel and high level waste is stored in dry conditions in interim storage facilities until a final repository will be ready for operation in the future. Dual purpose casks are used for the packaging of the materials to ensure transportability at any time during long-term interim storage and fulfilling both requirements from interim storage as well as transport. Germany’s 2011 decision to phase out utilization of nuclear energy has led to a high demand for casks for decommissioning power plants as quickly as possible. Cask serial production will continue at a high level for the next few years. In the process, a system of quality assurance measures ensures that every cask reliably complies with design specification according the transport approval certificate and storage license. &#13;
When the first casks were manufactured in Germany about 40 years ago, quality assurance was oriented towards pressure vessel and general nuclear manufacturing. The system was continuously adapted to the state of the art. In particular serial production, globalization of manufacturing and deep supplier chains had to be taken into account. The system of quality assurance measures applied today for cask manufacturing shall be presented here.&#13;
The quality assurance measures can be categorized as system-related and design-related. With respect to transport regulations for example, they are independently approved and monitored by the competent authority BAM and its authorized inspection experts. The qualification of the applicant is reviewed first in the context of the design approval procedure and then periodically. The qualification of component and cask manufacturers is also checked at the start of manufacturing and then periodically. Before the start of manufacturing documents such as quality plans are pre-assessed. Manufacturing and inspection are then performed according to the pre-assessed documents. Cask manufacturing is completed by a final inspection before commissioning. Periodic inspections during operation ensure preservation of the properties specified in the approval certificate over package life time. The type and the time interval of periodic inspections depend on the use of the package. For transport after interim storage a system of specific tests and inspections is prescribed.</abstract>
    <enrichment key="eventName">20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 22)</enrichment>
    <enrichment key="eventPlace">Juan-les-Pins, France</enrichment>
    <enrichment key="eventStart">11.06.2023</enrichment>
    <enrichment key="eventEnd">15.06.2023</enrichment>
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    <author>O. Schilling</author>
    <author>Martin Neumann</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Radioactive material</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Quality Assurance Measures</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Dual Purpose Casks</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Independent Surveillance</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="institutes" number="">3.3 Sicherheit von Transportbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
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