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    <title language="eng">Drop tests assessment of internal shock absorbers for packages loaded with encapsulations for damaged spent nuclear fuel</title>
    <abstract language="eng">Damaged spent nuclear fuel (DSNF) can be loaded in German dual-purpose casks (DPC) for transport and interim storage. Encapsulations are needed to guarantee a safe handling and a tight closure, separated from the package enclosure. These encapsulations shall be durable and leak-tight for a long storage period, because they are usually not accessible within periodical inspections of the DPC. Due to the general design of DPCs for standard fuel assemblies, specific requirements have to be considered for the design of encapsulations for DSNF to ensure the loading in existing package designs. Especially the primary lid system of a DPC is designed for maximum loads due to the internal impact of the content during drop test conditions. The main difference of encapsulations for damaged spent nuclear fuel is that they have usually a much higher stiffness than standard fuel assemblies. Therefore the design of an internal shock absorber, e.g. at the head of an encapsulation is required to reduce mechanical loads to the primary lid system during impacts.&#13;
BAM as part of the German competent authority system is responsible for the safety assessment of the mechanical and thermal package design, the release of radioactive material and the quality assurance of package manufacturing and operation. Concerning the mechanical design of the encapsulation BAM was involved in the comprehensive assessment procedure during the package design approval process. An internal shock absorber was developed by the package designer with numerical analyses and experimental drop tests. Experimental drop tests are needed to cover limiting parameters regarding, e.g. temperature and wall thickness of the shock absorbing element to enable a detailed specification of the whole load-deformation behavior of the encapsulation shock absorber.&#13;
The paper gives an overview of the assessment work by BAM and points out the main findings which are relevant for an acceptable design of internal shock absorbers. The physical drop tests were planned on the basis of pre-investigations of the applicant concerning shape, dimension and material properties. In advance of the final drop tests the possible internal impact behavior had to be analyzed and the setup of the test facility had to be validated. The planning, performance and evaluation of the final drop tests were witnessed and assessed by BAM. In conclusion it could be approved that the German encapsulation system for damaged spent nuclear fuel with shock absorbing components can be handled similar to standard fuel assemblies in existing package designs.</abstract>
    <parentTitle language="eng">Pressure Vessels &amp; Piping Conference 2020</parentTitle>
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    <author>Lars Müller</author>
    <author>Robert Scheidemann</author>
    <author>Thorsten Schönfelder</author>
    <author>Steffen Komann</author>
    <author>Frank Wille</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Encapsulations for damaged spent nuclear fuel</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Drop tests</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Internal shock absorber</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Design assessment of RAM packages</value>
    </subject>
    <collection role="ddc" number="611">Menschliche Anatomie, Zytologie, Histologie</collection>
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  <doc>
    <id>51550</id>
    <completedYear/>
    <publishedYear>2020</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst/>
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    <title language="eng">Drop tests assessment of internal shock absorbers for packages loaded with encapsulations for damaged spent nuclear fuel</title>
    <abstract language="eng">Damaged spent nuclear fuel (DSNF) can be loaded in German dual-purpose casks (DPC) for transport and interim storage. Encapsulations are needed to guarantee a safe handling and a tight closure, separated from the package enclosure. These encapsulations shall be durable and leak-tight for a long storage period, because they are usually not accessible within periodical inspections of the DPC. Due to the general design of DPCs for standard fuel assemblies, specific requirements have to be considered for the design of encapsulations for DSNF to ensure the loading in existing package designs. Especially the primary lid system of a DPC is designed for maximum loads due to the internal impact of the content during drop test conditions. The main difference of encapsulations for damaged spent nuclear fuel is that they have usually a much higher stiffness than standard fuel assemblies. Therefore the design of an internal shock absorber, e.g. at the head of an encapsulation is required to reduce mechanical loads to the primary lid system during impacts.&#13;
BAM as part of the German competent authority system is responsible for the safety assessment of the mechanical and thermal package design, the release of radioactive material and the quality assurance of package manufacturing and operation. Concerning the mechanical design of the encapsulation BAM was involved in the comprehensive assessment procedure during the package design approval process. An internal shock absorber was developed by the package designer with numerical analyses and experimental drop tests. Experimental drop tests are needed to cover limiting parameters regarding, e.g. temperature and wall thickness of the shock absorbing element to enable a detailed specification of the whole load-deformation behavior of the encapsulation shock absorber.&#13;
The paper gives an overview of the assessment work by BAM and points out the main findings which are relevant for an acceptable design of internal shock absorbers. The physical drop tests were planned on the basis of pre-investigations of the applicant concerning shape, dimension and material properties. In advance of the final drop tests the possible internal impact behavior had to be analyzed and the setup of the test facility had to be validated. The planning, performance and evaluation of the final drop tests were witnessed and assessed by BAM. In conclusion it could be approved that the German encapsulation system for damaged spent nuclear fuel with shock absorbing components can be handled similar to standard fuel assemblies in existing package designs.</abstract>
    <enrichment key="eventName">Pressure Vessels &amp; Piping Conference 2020</enrichment>
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    <author>Lars Müller</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Drop tests</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Internal shock absorber</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Design assessment of RAM packages</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Encapsulations for damaged 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="">3.3 Sicherheit von Transportbehältern</collection>
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  </doc>
  <doc>
    <id>61351</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
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    <edition/>
    <issue/>
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    <publisherPlace>New York</publisherPlace>
    <creatingCorporation>The American Society of Mechanical Engineers (ASME)</creatingCorporation>
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    <title language="eng">Tightness Verification Procedures of Welded Lids for Encapsulations for Damaged Spent Nuclear Fuel in the Design Approval Process of Dual-Purpose Cask (DPC)</title>
    <abstract language="eng">Damaged spent nuclear fuel (DSNF) can be loaded in Germany in dual-purpose casks (DPC) for transport and interim storage. Therefore, special encapsulations are needed to guarantee a safe handling and a reliable tight closure. These encapsulations must be durable and leak-tight for a possible long interim storage term and the transport thereafter. Because components inside the package containment like these special encapsulations are not accessible within periodical inspections of the DPC.&#13;
Due to the general design of DPCs for standard fuel assemblies, specific requirements needed to be considered for the design of the special encapsulations for DSNF to ensure the loading in existing package designs. Especially the content of the encapsulations with damaged spent nuclear fuel must be avoided to enter in the inner space of the DPC, due to the possible higher amount of corrosive and radioactive material as well as greater quantities of water, which can be trapped in defect fuel rods. The main difference of the encapsulations for damaged spent nuclear fuel is that they have no replaceable sealing as tightness barrier. They have a specific welded lid. Therefor the tightness needs to be shown in the design approval process for all transport conditions especially for accident conditions, where high internal impact loads can occur.&#13;
As German competent authority, BAM is responsible for the safety assessment of the mechanical and thermal design, the release of radioactive material and the quality assurance aspects of manufacturing and operation. Concerning the mechanical package design BAM carried out a comprehensive safety assessment for the specific encapsulations during the package design approval process of the commonly used DPC. For the proof of tightness of the welded lid no representative standards are available. For minor loads the German KTA standards can verify the tightness of a weld seam but only for low stresses. For higher impact loads physical tests are required. These tests were performed with representative drop tests and highly sensitive leak tests.&#13;
The present paper gives an overview of the applied assessments by BAM and points out the main findings of the mechanical design for the welded sealing. Both verification procedures are described, on one hand with German standards for lower loads and on the other hand with physical tests for higher impact loads. The drop tests based on complex numerical pre-calculations by the applicant which result in optimized dimensions and properties of the test specimens and representative drop heights for the drop tests. The horizontal drop orientation with a plane-parallel impact on the unyielding basement was most challenging and led to specific adaptations during preparation and conduction of the drop tests performed by BAM. With leak tests and the corresponding test evaluation, it could be finally approved that the specific welded sealing of an encapsulations for DSNF is tight for all transport conditions.</abstract>
    <parentTitle language="eng">Proceedings of the ASME 2024 Pressure Vessels &amp; Piping Conference</parentTitle>
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    <author>Lars Müller</author>
    <author>Robert Scheidemann</author>
    <author>Tino Neumeyer</author>
    <author>Steffen Komann</author>
    <author>Frank Wille</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Tightness verification</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Weld seams</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Drop tests</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Encapsulations for damaged spent nuclear fuel</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Helium leak tests</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
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  </doc>
  <doc>
    <id>57770</id>
    <completedYear/>
    <publishedYear>2023</publishedYear>
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    <language>eng</language>
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    <publisherName>World Nuclear Transport Institute (WNTI)</publisherName>
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    <title language="eng">Introduction of the German ageing management guide for packages for transport of radioactive materials</title>
    <abstract language="eng">The consideration of ageing mechanisms is with integration of the new para 613A into IAEA SSR-6 (Rev. 1) now obligatory for the design of transport packages. In addition, para 809(f) requires for packages intended to be used for shipment after storage the consideration of the effects of ageing mechanisms during storage in safety analyses and the implementation of corresponding instructions for operation and maintenance. Para 503(e) requires that all packaging components and radioactive contents have been maintained during storage in a manner that all requirements specified in IAEA SSR-6 (Rev.1) and in the applicable certificates of approval have been fulfilled.&#13;
&#13;
The evaluation of ageing mechanisms and their effects including monitoring are part of BAM’s authority assessment tasks related to the mechanical and thermal package design and quality assurance aspects. BAM has compiled a guideline for the implementation of ageing assessment and of the measures for ageing management of the approval procedure based on requirements of IAEA SSR-6 (Rev.1). The guideline is applicable only for packages requiring a competent authority approval.&#13;
&#13;
The paper aims to describe the structure of the guideline and the general approach for ageing management requirements. The type and amount of measures for ageing management depend mainly on the use of the package and on the ageing effects for the component, which result from relevant ageing mechanisms during package operation time.&#13;
&#13;
The implementation of measures for ageing management is divided into three levels – systemic measures, package design related measures and documentation. The systemic measures are attributed to the general management system and define the whole activities for organization of ageing management like structure, responsibilities, documentation, reports and evaluation. The package design related measures are defined in an ageing management plan (AMP). These measures shall ensure that the anticipated changes of the package design under consideration of ageing effects still complies with the design approval specification. Therefore, an ageing surveillance program (ASP) and, if necessary, a gap analysis program shall be developed. The ageing management documentation (AMD) ensures the continuous documentation of the compliance of a specific package to the approved package design, comprising mainly records resulting from operation and surveillance.</abstract>
    <parentTitle language="eng">Proceedings of the 20th international symposium on the packaging and transportation of radioactive materials</parentTitle>
    <enrichment key="eventName">PATRAM22 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials</enrichment>
    <enrichment key="eventPlace">Juan-les-Pins, France</enrichment>
    <enrichment key="eventStart">11.06.2023</enrichment>
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    <enrichment key="opus.source">publish</enrichment>
    <author>Sven Schubert</author>
    <author>Adrian Reichardt</author>
    <author>Lars Müller</author>
    <author>Martin Neumann</author>
    <author>Steffen Komann</author>
    <author>Frank Wille</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Guide</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ageing</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Mechanism</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Package</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Management</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
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    <collection role="institutes" number="">3.3 Sicherheit von Transportbehältern</collection>
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  </doc>
  <doc>
    <id>64023</id>
    <completedYear/>
    <publishedYear>2025</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>11</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume>2025</volume>
    <type>conferenceobject</type>
    <publisherName>Institute of Nuclear Materials Management (INMM)</publisherName>
    <publisherPlace>Indianapolis</publisherPlace>
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    <title language="eng">Thightness Assessment of welded Lids for Encapsulations of damaged Spent Nuclear Fuel in the Design Approval Process of Dual-Purpose Casks (DPC)</title>
    <abstract language="eng">The disposal of spent nuclear fuel in Germany is ensured using dual-purpose casks (DPC) for transport and interim storage. The leak tightness of the DPC and resulting containment is one of the most important aspects. Additional encapsulations are required for damaged spent nuclear fuel (DSNF) to guarantee safe handling and a separate tight closure. Due to the general design of DPCs for standard fuel assemblies should special requirements be considered for the design of the encapsulations for DSNF to ensure the loading in existing package designs. The absence of a replaceable sealing in the tightness barrier is the main difference for the encapsulations for damaged spent nuclear fuel. Instead, they are welded shut with a lid. The leak tightness of the encapsulation shall be proven in the design approval process for all transport conditions. This is especially valid for accident conditions of transport, where high internal impact forces may occur.&#13;
BAM as German competent authority is responsible for the safety assessment of mechanical and thermal design, retention of radioactive material and quality assurance aspects of manufacturing and operation. BAM carried out a comprehensive safety assessment concerning the mechanical package design. As there are no representative standards for verifying the leak tightness of a welded lid, two approaches were being pursued. Established German standards may be used for verifying the leak tightness of a weld - but limited to low stresses. Therefore, physical tests were required for higher impact loads. Representative drop tests and highly sensitive leakage tests were performed.&#13;
The paper presents an overview of the containment assessment by BAM and points out the main findings for the design of welds regarding leak tightness. Both verification procedures are described, on one hand with German standards for lower loads and on the other hand with physical tests for higher impact loads. The leak tightness of the encapsulation weld could be approved based on leakage tests and a corresponding evaluation for all transport conditions.</abstract>
    <parentTitle language="eng">Proceedings of the PATRAM 2025 Conference</parentTitle>
    <enrichment key="eventName">PATRAM 2025</enrichment>
    <enrichment key="eventPlace">San Antonio, Texas, USA</enrichment>
    <enrichment key="eventStart">27.07.2025</enrichment>
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    <author>Lars Müller</author>
    <author>Robert Scheidemann</author>
    <author>Tino Neumeyer</author>
    <author>Martin Neumann</author>
    <author>Steffen Komann</author>
    <author>Frank Wille</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Transport packages</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Tightness assessment</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Welded lids</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Encapsulations</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</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="">Graue Literatur</collection>
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  </doc>
  <doc>
    <id>64026</id>
    <completedYear/>
    <publishedYear>2025</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>8</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume>2025</volume>
    <type>conferenceobject</type>
    <publisherName>Institute of Nuclear Materials Management (INMM)</publisherName>
    <publisherPlace>Indianapolis</publisherPlace>
    <creatingCorporation/>
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    <title language="eng">Multilateral approval of a SCO-III (three spent steam generator lower parts from EDF Fessenheim nuclear power plant) multimodal shipment</title>
    <abstract language="eng">A Surface Contaminated Object (SCO) is defined as a solid object which is not itself radioactive, but which has radioactive material distributed on its surfaces. The 2018 Edition of the IAEA Regulations for the Safe Transport of Radioactive Material (SSR-6 Rev. 1) introduced the group SCO-III to transport unpackaged large objects.&#13;
The two pressurized water reactors of the Fessenheim Nuclear Power Plant, owned by EDF, were shut down in 2020. Previously, for maintenance operations in the year 2002 and 2010, the three steam generators – over 20 metres long, 300 tons – of each reactor were taken out from the reactors, divided in two and stored on site:&#13;
- EDF shipped the six upper parts, as SCO-I, to Cyclife, in Studsvik (Sweden), in 2021, to be&#13;
recycled;&#13;
- EDF plans two transports of the lower parts (three parts for each transport), as SCO-III&#13;
(considering the higher activity due to the NPP primary coolant) to the same facility. Modes of&#13;
transport are, from consignor to consignee, road, inland waterways and maritime. The countries crossed by inland waterways, from Neuf-Brisach to Dunkerque (France), are France, Germany, Belgium and the Netherlands.&#13;
According to the para. 825 of SSR-6 Rev. 1, “multilateral approval shall be required for the shipment of SCO-III”. EDF has submitted to all Competent Authorities an application (para. 827A of SSR-6 Rev. 1), including a transport plan and demonstrations to withstand normal conditions of transport.&#13;
Despite SSR-6 Rev. 1, national regulations have their own specificities and each Competent Authority has its own assessment procedures. Nevertheless, involved Authorities agreed for a joint review, which included:&#13;
- the understanding of national regulatory requirements,&#13;
- the expected level of detail and quality of the safety case,&#13;
- the feedback from previous shipments of similar type,&#13;
- the contents, including the conditions, of the approvals to be issued.&#13;
This joint review demonstrated good coordination between authorities, enabling the application to be processed rapidly, and the various authorities to benefit from each other's expertise and feedback. This was of particular importance, as it was the first application for the transport of an SCO-III assessed by the Competent Authorities involved.&#13;
Finally, the authorities were able to issue approvals, whose format is certainly in line with their national provisions, but whose content and level of requirements in terms of transport safety are consistent.</abstract>
    <parentTitle language="eng">Proceedings of the PATRAM 2025 Conference</parentTitle>
    <enrichment key="eventName">PATRAM 2025</enrichment>
    <enrichment key="eventPlace">San Antonio, Texas, USA</enrichment>
    <enrichment key="eventStart">27.07.2025</enrichment>
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    <enrichment key="opus.source">publish</enrichment>
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    <author>Matthieu Guitton</author>
    <author>Thierry Chrupek</author>
    <author>Vincent Leblanc</author>
    <author>Jeroen Van derTuijn</author>
    <author>Dominik Kaufhold</author>
    <author>Frederik Kesting</author>
    <author>Lars Müller</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Shipment</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Multilateral approval</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Steam generator</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>SCO-III</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</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="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>64017</id>
    <completedYear/>
    <publishedYear>2025</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst/>
    <pageLast/>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>lecture</type>
    <publisherName/>
    <publisherPlace/>
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    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Thightness Assessment of welded Lids for Encapsulations of damaged Spent Nuclear Fuel in the Design Approval Process of Dual-Purpose Casks (DPC)</title>
    <abstract language="eng">The disposal of spent nuclear fuel in Germany is ensured using dual-purpose casks (DPC) for transport and interim storage. The leak tightness of the DPC and resulting containment is one of the most important aspects. Additional encapsulations are required for damaged spent nuclear fuel (DSNF) to guarantee safe handling and a separate tight closure. Due to the general design of DPCs for standard fuel assemblies should special requirements be considered for the design of the encapsulations for DSNF to ensure the loading in existing package designs. The absence of a replaceable sealing in the tightness barrier is the main difference for the encapsulations for damaged spent nuclear fuel. Instead, they are welded shut with a lid. The leak tightness of the encapsulation shall be proven in the design approval process for all transport conditions. This is especially valid for accident conditions of transport, where high internal impact forces may occur.&#13;
BAM as German competent authority is responsible for the safety assessment of mechanical and thermal design, retention of radioactive material and quality assurance aspects of manufacturing and operation. BAM carried out a comprehensive safety assessment concerning the mechanical package design. As there are no representative standards for verifying the leak tightness of a welded lid, two approaches were being pursued. Established German standards may be used for verifying the leak tightness of a weld - but limited to low stresses. Therefore, physical tests were required for higher impact loads. Representative drop tests and highly sensitive leakage tests were performed.&#13;
The paper presents an overview of the containment assessment by BAM and points out the main findings for the design of welds regarding leak tightness. Both verification procedures are described, on one hand with German standards for lower loads and on the other hand with physical tests for higher impact loads. The leak tightness of the encapsulation weld could be approved based on leakage tests and a corresponding evaluation for all transport conditions.</abstract>
    <enrichment key="eventName">PATRAM 2025</enrichment>
    <enrichment key="eventPlace">San Antonio, TX, USA</enrichment>
    <enrichment key="eventStart">27.07.2025</enrichment>
    <enrichment key="eventEnd">01.08.2025</enrichment>
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    <author>Lars Müller</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Tightness assessment</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Welded lids</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Transport packages</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Encapsulations</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.3 Sicherheit von Transportbehältern</collection>
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  </doc>
  <doc>
    <id>63878</id>
    <completedYear/>
    <publishedYear>2025</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>10</pageLast>
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    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Shipment of SCO-II – Authority assessment of mechanical aspects and quality management</title>
    <abstract language="eng">The decommissioning of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, or coolant pumps. These components or objects are large in size and mass with up to 6 meters in diameter and 20 meters in length and a weight of up to 400 tons. They are often transported to a storage, disposal, or recycling facility. Large components from nuclear facilities may often not be packed and need to be transported unpackaged due to size and weight.</abstract>
    <parentTitle language="eng">Proceedings of the PATRAM 2025 Conference</parentTitle>
    <enrichment key="eventName">PATRAM 2025</enrichment>
    <enrichment key="eventPlace">San Antonio, TX, USA</enrichment>
    <enrichment key="eventStart">27.07.2025</enrichment>
    <enrichment key="eventEnd">01.08.2025</enrichment>
    <enrichment key="opus.source">publish</enrichment>
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    <author>Steffen Komann</author>
    <author>Lars Müller</author>
    <author>Martin Neumann</author>
    <author>Frank Wille</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>radioaktive Stoffe, Rückbau kerntechnischer Anlagen, Transport, Gefahrgut</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.3 Sicherheit von Transportbehältern</collection>
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  </doc>
  <doc>
    <id>60186</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>8</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume>2024</volume>
    <type>conferenceobject</type>
    <publisherName>Nuclear Institute</publisherName>
    <publisherPlace>London</publisherPlace>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Ageing management measures of transport packages for radioactive materials</title>
    <abstract language="eng">The consideration of ageing mechanisms is with integration into the IAEA-requirements now obligatory for the design of transport packages. For packages intended to be used for shipment after storage the consideration of the effects of ageing mechanisms during storage in safety analyses and the implementation of corresponding instructions for operation and maintenance are required. Therefore, these packages have been maintained during storage in a manner that all requirements specified in the IAEA-requirements and in the applicable certificates of approval have been fulfilled.&#13;
BAM is the competent authority in Germany for evaluating the thermal and mechanical design of packages for the transport of radioactive materials. In this context, BAM also evaluates the aging mechanisms and aging effects occurring during the long-term storage period of such transport packages. BAM has developed a guideline for the implementation of ageing assessment and of the measures for ageing management into the approval procedure based on IAEA-requirements.&#13;
The goal of the presented paper is to introduce this guideline and the general approach for ageing management requirements and to describe the application to a specific package design. The type and amount of the ageing management not only depend on the use of the packaging, but also on the accessibility of the components that are considered in the ageing assessment. The different operating phases of a packaging are considered for the ageing assessment, such as loading, interim storage or transport.&#13;
The implementation of measures for ageing management is divided into three levels – systemic measures, package design related measures and documentation. The systemic measures are attributed to the general management system and define the whole activities for organization of ageing management like structure, responsibilities, documentation, reports and evaluation. The package design related measures are defined in an ageing management plan. These measures shall ensure that the anticipated changes of the package design under consideration of ageing effects still complies with the design approval specification. Therefore, an ageing surveillance program and, if necessary, a gap analysis program shall be developed. The ageing management documentation ensures the continuous documentation of the compliance of a specific package to the approved package design, comprising mainly records resulting from operation and surveillance.</abstract>
    <parentTitle language="eng">Proceedings of Conference RAMTRANS 2024</parentTitle>
    <enrichment key="eventName">RAMTRANS 2024</enrichment>
    <enrichment key="eventPlace">London,  United Kingdom</enrichment>
    <enrichment key="eventStart">15.05.2024</enrichment>
    <enrichment key="eventEnd">17.05.2024</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <author>Sven Schubert</author>
    <author>Steffen Komann</author>
    <author>Martin Neumann</author>
    <author>Adrian Reichardt</author>
    <author>Lars Müller</author>
    <author>Frank Wille</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Radioactive Materials</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ageing Management</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Transport Packages</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Dual Purpose Casks</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</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="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
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