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  <doc>
    <id>50987</id>
    <completedYear/>
    <publishedYear>2020</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">Retaining Competence for Safe Transports of Radioactive Material</title>
    <abstract language="eng">The nuclear phase out in Germany will lead to a challenge finding skilled staff for all areas in nuclear business. Transport of spent fuel and high-level nuclear waste is an ongoing task over the next decades during the decommissioning of nuclear power plants and after the interim storage. The international regulations were adapted and address the issue of transports after storage now. The topic of aging and assessment of the development of the state-of-art technology are in focus. German research funds are supporting the investigations of future challenges of transport and storage of the nuclear heritage. BAM is involved and works on several topics related to the aging issue of packages. The behaviour of spent fuel claddings, long-term behaviour of metal and elastomeric gaskets and sensor/monitoring techniques of waste disposal are examples of research fields.</abstract>
    <enrichment key="eventName">7th Nuclear Decommissioning &amp; Waste Management Summit 2020</enrichment>
    <enrichment key="eventPlace">London, UK</enrichment>
    <enrichment key="eventStart">12.02.2020</enrichment>
    <enrichment key="eventEnd">13.02.2020</enrichment>
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    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <enrichment key="InvitedTalks">0</enrichment>
    <author>Frank Wille</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Aging management</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Transport</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent fuel behaviour</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Disposal</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>50428</id>
    <completedYear/>
    <publishedYear>2020</publishedYear>
    <thesisYearAccepted/>
    <language>deu</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="deu">Kerntechnische Entsorgung Herausforderungen im Transport und bei der Lagerung</title>
    <abstract language="deu">Nach der Reaktorkatastrophe in Fukushima 2011 hat die Bundesregierung beschlossen die Nutzung der Kernenergie bis zum Jahr 2022 zu beenden. Seitdem erfolgt der Rückbau und die Stilllegung der kerntechnischen Anlagen in Deutschland. Dieser Vortrag erörtert die Herausforderungen, die sich mit dem Transport und der Lagerung von Kernbrennstoffen ergeben und leitet daraus Anforderungen ab, die an Transportbehälter gestellt werden.</abstract>
    <enrichment key="eventName">Ausgewählte Kapitel der Werkstofftechnik - Brand- und Explosionsverhalten von Werkstoffen</enrichment>
    <enrichment key="eventPlace">Wuppertal, Germany</enrichment>
    <enrichment key="eventStart">24.01.2020</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <enrichment key="InvitedTalks">0</enrichment>
    <author>Thorsten Schönfelder</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Kerntechnische Entsorgung</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Rückbau</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Transport- und Lagerbehälter</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>50734</id>
    <completedYear/>
    <publishedYear>2020</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1517</pageFirst>
    <pageLast>1543</pageLast>
    <pageNumber/>
    <edition/>
    <issue>3</issue>
    <volume>28</volume>
    <type>article</type>
    <publisherName>Springer Nature</publisherName>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>1</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">High-Order Accurate Methods for the Numerical Analysis of a Levitation Device</title>
    <abstract language="eng">This paper establishes different axisymmetric and two-dimensional models for a levitation device. Therein, the Maxwell equations are combined with the balance of linear momentum. Different possible formulations to describe the Maxwell equations are presented and compared and discussed in the example. A high order finite element discretization using Galerkin’s method in space and the generalized Newmark−alpha method in time are developed for the electro-magneto-mechanical approach. Several studies on spatial and temporal discretization with respect to convergence will be investigated. In addition, the boundary influences and the domain size with respect to the levitation device are also examined.</abstract>
    <parentTitle language="eng">Archives of Computational Methods in Engineering</parentTitle>
    <identifier type="doi">10.1007/s11831-020-09427-z</identifier>
    <identifier type="issn">1134-3060</identifier>
    <enrichment key="date_peer_review">07.05.2020</enrichment>
    <author>Tobias Gleim</author>
    <author>Hefeng Chen</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Maxwell equations</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Levitation</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>High-Order</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Accurate-Numerical-Methods</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Generalized Newmark-alpha</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="literaturgattung" number="">Verlagsliteratur</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>51103</id>
    <completedYear/>
    <publishedYear>2020</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>7</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume>2020</volume>
    <type>conferenceobject</type>
    <publisherName>ASME</publisherName>
    <publisherPlace>New York</publisherPlace>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>1</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Mechanical and thermal assessment by BAM of a new package design for the transport of SNF from a german research reactor</title>
    <abstract language="eng">For disposal of the research reactor of the Technical University Munich FRM II a new transport and storage cask design was under approval assessment by the German authorities on the basis of International Atomic Energy Agency (IAEA) requirements. The cask body is made of ductile cast iron and closed by two bolted lid systems with metal seals. The material of the lids is stainless steel. On each end of the cask the wood-filled impact limiters are installed to reduce impact loads to the cask under drop test conditions. In the cavity of the cask a basket for five spent fuel elements is arranged. &#13;
This design has been assessed by the Bundesanstalt für Materialforschung und -prüfung (BAM) in view to the mechanical and thermal safety analyses, the activity release approaches, and subjects of quality assurance and surveillance for manufacturing and operation of the package. &#13;
For the mechanical safety analyses of the package a combination of experimental testing and analytical/numerical calculations were applied. In total, four drop tests were carried out at the BAM large drop test facility. Two tests were carried out as a full IAEA drop test sequence consisting of a 9m drop test onto an unyielding target and a 1m puncture bar drop test. The other two drop tests were performed as single 9m drop tests and completed by additional analyses for considering the effects of an IAEA drop test sequence.&#13;
The main objectives of the drop tests were the investigation of the integrity of the package and its safety against release of radioactive material as well as the test of the fastening system of the impact limiters. Furthermore, the acceleration and strain signals measured during the tests were used for the verification of finite-element (FE) models applied in the safety analysis of the package design.</abstract>
    <parentTitle language="eng">Pressure Vessels  Piping Conference</parentTitle>
    <enrichment key="eventName">Pressure Vessels &amp; Piping Conference 2020</enrichment>
    <enrichment key="eventPlace">Online meeting</enrichment>
    <enrichment key="eventStart">03.08.2020</enrichment>
    <enrichment key="eventEnd">07.08.2020</enrichment>
    <enrichment key="date_peer_review">19.08.2024</enrichment>
    <author>Steffen Komann</author>
    <author>Viktor Ballheimer</author>
    <author>Thomas Quercetti</author>
    <author>Robert Scheidemann</author>
    <author>Frank Wille</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Mechanik</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>radioaktives Material</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Transportbehälter</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Antragsverfahren</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Zulassungen</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Typ-B Versandstück</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Thermik</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>51550</id>
    <completedYear/>
    <publishedYear>2020</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">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>
    <enrichment key="eventPlace">Online meeting</enrichment>
    <enrichment key="eventStart">03.08.2020</enrichment>
    <enrichment key="eventEnd">07.08.2020</enrichment>
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    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <enrichment key="InvitedTalks">0</enrichment>
    <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>
    <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>51192</id>
    <completedYear/>
    <publishedYear>2020</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>9</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName>ASME</publisherName>
    <publisherPlace>New York</publisherPlace>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>1</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Outcomes of three large scale fire reference tests conducted in BAM fire test facility</title>
    <abstract language="eng">Packages for the transport of high-level radioactive material are designed to withstand severe accidents. Hypothetical severe accident conditions are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials.&#13;
One of these accident conditions is the thermal test, mainly consisting of a 30 minute fully engulfing 800°C pool fire or an equally severe fire test. The heat fluxes into the package depend substantially on the fire characteristics and the Surface temperature of the package.&#13;
Fire tests can be performed at BAM on a propane gas fire test facility. In order to investigate the heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for multiple use. The package represented the outer geometry of a specific transport cask for radioactive waste. The fire reference package is a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 1500 mm and a diameter of 1050 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material.&#13;
Three open-air fire tests were performed in the BAM propane gas fire test facility. The flames exposure time period varied slightly for the fire tests. The wind direction as well as the wind Speed were measured and changed between and during the tests. Test stand parameters such as wind shield location and propane gas volume flow were chosen constant for the three tests. The locally measured fire reference package steel sheet temperatures were used for the calculation of heat fluxes as function of time and surface temperature. The measured temperatures allowed further calculations. In a first approach effective fire characteristics of the propane gas fire, including the flame temperature, the fire convection coefficient and a Radiation exchange coefficient mathematically describing the determined average heat flux over the surface temperature were calculated.</abstract>
    <parentTitle language="eng">Pressure Vessels &amp; Piping Conference 2020</parentTitle>
    <enrichment key="eventName">Pressure Vessels &amp; Piping Conference 2020</enrichment>
    <enrichment key="eventPlace">Online meeting</enrichment>
    <enrichment key="eventStart">03.08.2020</enrichment>
    <enrichment key="eventEnd">07.08.2020</enrichment>
    <enrichment key="date_peer_review">19.08.2024</enrichment>
    <author>Martin Feldkamp</author>
    <author>Thomas Quercetti</author>
    <author>Frank Wille</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Fire</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Testing</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Large scale testing</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Calorimeter</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Heat flux</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>51194</id>
    <completedYear/>
    <publishedYear>2020</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">Outcomes of three large scale fire reference tests conducted in BAM fire test facility</title>
    <abstract language="eng">Packages for the transport of high-level radioactive material are designed to withstand severe accidents. Hypothetical severe accident conditions are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials.&#13;
One of these accident conditions is the thermal test, mainly consisting of a 30 minute fully engulfing 800°C pool fire or an equally severe fire test. The heat fluxes into the package depend substantially on the fire characteristics and the Surface temperature of the package.&#13;
Fire tests can be performed at BAM on a propane gas fire test facility. In order to investigate the heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for multiple use. The package represented the outer geometry of a specific transport cask for radioactive waste. The fire reference package is a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 1500 mm and a diameter of 1050 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material.&#13;
Three open-air fire tests were performed in the BAM propane gas fire test facility. The flames exposure time period varied slightly for the fire tests. The wind direction as well as the wind Speed were measured and changed between and during the tests. Test stand parameters such as wind shield location and propane gas volume flow were chosen constant for the three tests. The locally measured fire reference package steel sheet temperatures were used for the calculation of heat fluxes as function of time and surface temperature. The measured temperatures allowed further calculations. In a first approach effective fire characteristics of the propane gas fire, including the flame temperature, the fire convection coefficient and a Radiation exchange coefficient mathematically describing the determined average heat flux over the surface temperature were calculated.</abstract>
    <enrichment key="eventName">Pressure Vessels &amp; Piping Virtual Conference 2020</enrichment>
    <enrichment key="eventPlace">Online meeting</enrichment>
    <enrichment key="eventStart">03.08.2020</enrichment>
    <enrichment key="eventEnd">07.08.2020</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <enrichment key="InvitedTalks">0</enrichment>
    <author>Martin Feldkamp</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Fire</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Testing</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Convection coefficient</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Large scale fire testing</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Propane gas fire test facility</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>51718</id>
    <completedYear/>
    <publishedYear>2020</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">Assessment of ageing effects at BAM</title>
    <abstract language="eng">In Germany, spent nuclear fuel and vitrified high active waste is stored in dual purpose casks (DPC) at interim storage facilities. In order to ensure the transportability of the DPC to a final repository in future, the maintenance of the package design approval is realized. Therefore, the assessment of possible ageing effects during interim storage is necessary to ensure an evaluation of the transportability. The lecture presents BAM's current ageing evaluation concept in the field of trasnport law.</abstract>
    <enrichment key="eventName">IRSN-BAM Symposium - Safety of Transport and Storage Packages</enrichment>
    <enrichment key="eventPlace">Online meeting</enrichment>
    <enrichment key="eventStart">19.11.2020</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <enrichment key="InvitedTalks">0</enrichment>
    <author>Adrian Reichardt</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ageing</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>DPC</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Interim storage</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ageing management</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>51778</id>
    <completedYear/>
    <publishedYear>2020</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">Outcomes of Three Large-Scale Fire Reference Tests Conducted in Propane Gas Fire Test Facility</title>
    <abstract language="eng">Packages for the transport of high-level radioactive material are designed to withstand severe accidents. Hypothetical severe accident conditions are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials. One of these accident conditions is the thermal test, mainly consisting of a 30 minute fully engulfing 800°C pool fire or an equally severe fire test. The heat fluxes into the package depend substantially on the fire characteristics and the surface temperature of the package. Fire tests can be performed at BAM on a propane gas fire test facility. In order to investigate the heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for multiple use. The package represented the outer geometry of a specific transport cask for radioactive waste. The fire reference package is a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 1500 mm and a diameter of 1050 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material. Three open-air fire tests were performed in the BAM propane gas fire test facility. The flames exposure time period varied slightly for the fire tests. The wind direction as well as the wind speed were measured and changed between and during the tests. Test stand parameters such as wind shield location and propane gas volume flow were chosen constant for the three tests. The locally measured fire reference package steel sheet temperatures were used for the calculation of heat fluxes as function of time and surface temperature. The measured temperatures allowed further calculations. In a first approach effective fire characteristics of the propane gas fire, including the flame temperature, the fire convection coefficient and a radiation exchange coefficient mathematically describing the determined average heat flux over the surface temperature were calculated.</abstract>
    <enrichment key="eventName">Abteilungskolloquium 3.3</enrichment>
    <enrichment key="eventPlace">Online meeting</enrichment>
    <enrichment key="eventStart">03.12.2020</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <enrichment key="InvitedTalks">0</enrichment>
    <author>Martin Feldkamp</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Fire</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Heat flux</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Convection coefficient</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Adiabatic surface temperature</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>IAEA</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>51686</id>
    <completedYear/>
    <publishedYear>2020</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">Fire reference package - Results of large-scale fire tests</title>
    <abstract language="eng">Packages for the transport of high-level radioactive material are designed to withstand severe accidents. Hypothetical severe accident conditions are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials. One of these accident conditions is the thermal test, mainly consisting of a 30 minute fully engulfing 800°C pool fire or an equally severe fire test. The heat fluxes into the package depend substantially on the fire characteristics and the surface temperature of the package. Fire tests can be performed at BAM on a propane gas fire test facility. In order to investigate the heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for multiple use. The package represented the outer geometry of a specific transport cask for radioactive waste. The fire reference package is a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 1500 mm and a diameter of 1050 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material. Three open-air fire tests were performed in the BAM propane gas fire test facility. The flames exposure time period varied slightly for the fire tests. The wind direction as well as the wind speed were measured and changed between and during the tests. Test stand parameters such as wind shield location and propane gas volume flow were chosen constant for the three tests. The locally measured fire reference package steel sheet temperatures were used for the calculation of heat fluxes as function of time and surface temperature. The measured temperatures allowed further calculations. In a first approach effective fire characteristics of the propane gas fire, including the flame temperature, the fire convection coefficient and a radiation exchange coefficient mathematically describing the determined average heat flux over the surface temperature were calculated.</abstract>
    <enrichment key="eventName">IRSN-BAM Symposium - Safety of Transport and Storage Packages</enrichment>
    <enrichment key="eventPlace">Online meeting</enrichment>
    <enrichment key="eventStart">19.11.2020</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">false</enrichment>
    <enrichment key="InvitedTalks">0</enrichment>
    <author>Martin Feldkamp</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Fire reference</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Calorimetric test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Propane</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Heat flux</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Propane gas fire test facility</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>52072</id>
    <completedYear/>
    <publishedYear>2021</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>e202000018</pageFirst>
    <pageLast/>
    <pageNumber/>
    <edition/>
    <issue>1</issue>
    <volume>20</volume>
    <type>conferenceobject</type>
    <publisherName>Wiley‐VCH GmbH</publisherName>
    <publisherPlace>Online</publisherPlace>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Numerical Analysis for an Electro‐Magneto‐Mechanical Phenomenon with High‐Order Accurate Methods</title>
    <abstract language="eng">This paper establishes an axisymmetric model for a levitation device. Therein, the Maxwell equations are combined with the balance of linear momentum. Different possible formulations to describe the MAXWELL equations are presented and compared and discussed in the example. A high order finite element discretization using GALERKIN's method in space and the generalized NEWMARK‐α method in time are developed for the electro‐magneto‐mechanical approach. Several studies on spatial and temporal discretization with respect to convergence will be investigated. In addition, the boundary influences and the domain size with respect to the levitation device are also examined.</abstract>
    <parentTitle language="eng">Proceedings in Applied Mathematics &amp; Mechanics</parentTitle>
    <identifier type="doi">10.1002/pamm.202000018</identifier>
    <identifier type="urn">urn:nbn:de:kobv:b43-520720</identifier>
    <enrichment key="eventName">International Association for Applied Mathematics and Mechanics</enrichment>
    <enrichment key="eventPlace">Kassel, Germany</enrichment>
    <enrichment key="eventStart">16.03.2020</enrichment>
    <enrichment key="eventEnd">20.03.2020</enrichment>
    <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-NC-ND - Namensnennung - Nicht kommerziell - Keine Bearbeitungen 4.0 International</licence>
    <author>Tobias Gleim</author>
    <author>Hefeng Chen</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>High-Order Methods</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Electro-Magneto-Mechanical</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Levitation Device</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="literaturgattung" number="">Verlagsliteratur</collection>
    <collection role="fulltextaccess" number="">Datei für die Öffentlichkeit verfügbar ("Open Access")</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
    <collection role="unnumberedseries" number="">Wissenschaftliche Artikel der BAM</collection>
    <thesisPublisher>Bundesanstalt für Materialforschung und -prüfung (BAM)</thesisPublisher>
    <file>https://opus4.kobv.de/opus4-bam/files/52072/gleim_pamm2020.pdf</file>
  </doc>
  <doc>
    <id>53242</id>
    <completedYear/>
    <publishedYear>2021</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">Fire Testing - Current Activities</title>
    <abstract language="eng">The presentation gives an overview over three current fire testing activities for testing transport packages for the transport of high-level radioactive material. Packages for the transport of high-level radioactive material must withstand severe hypothetical accidents. Regulatory test conditions shall cover these severe accident conditions and consist of mechanical tests and a following thermal test. To withstand the mechanical tests heavy weight packages are often designed with impact limiters consisting of wood encapsulated in steel sheets. The thermal test is defined precisely in the IAEA-regulations as a 30 minute fully engulfing 800 °C fire. After the fire phase a pre-damaged impact limiter might continue burning or smoldering and influence the cask thermal behavior with its energy release. The energy transferred from the impact limiter to the cask is of importance for the safety of transport packages. A full-scale fire test with an impact limiter of 2.3 m in diameter and filled with spruce wood was designed and performed. The impact limiter continued burning for 3 days. Energy transfer and temperature measurements were performed. A new test is designed to examine pressure build up and possible mechanical failures in an undamaged impact limiter with a diameter of 1.9 m. The test is designed to measure heat flux from the impact limiter in case of its ignition and burning. To furthermore examine the burning behavior of steel encapsulated wood piles, tests are prepared with wood fire containers and an infrared ignition source. A fire reference package for calorimetric tests was designed. The fire reference package design consists of a closed steel sheet cylinder with a length of 1,500 mm, an exterior diameter of 1,050 mm, and a wall thickness of 10 mm. The cylinder was filled with refractory insulation material and instrumented with thermocouples distributed all over the cylinder. The measured local steel sheet temperatures allow the determination of local as well as overall integral heat fluxes versus time and versus surface temperature. Currently, tests are planned with a new and relatively small fire reference package design. The new design will have a steel sheet cylinder with a length of 182 mm, an exterior diameter of 120 mm. Additionally heat flux sensors will be used in the new fire reference package test design. The propane gas fire will be adjusted with respect to the outcome of the fire tests to meet the requirements of the IAEA-fire.</abstract>
    <enrichment key="eventName">Technical Exchange IRSN – BAM Transport &amp; Storage of Packages for Radioactive Material</enrichment>
    <enrichment key="eventPlace">Berlin, Germany</enrichment>
    <enrichment key="eventStart">08.09.2021</enrichment>
    <enrichment key="eventEnd">09.08.2021</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>Martin Feldkamp</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Heat flux sensor</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Propane gas fire</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Wood fire</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>53251</id>
    <completedYear/>
    <publishedYear>2021</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">Prestressed bolt connections under lateral displacements</title>
    <abstract language="eng">The containment system of transport packages for spent nuclear fuel and high-level waste usually includes bolted lids with metal gaskets. The specified transport condition imply high loading on the lids and the bolt connections of the package. The response of the lid systems on these load conditions is generally investigated by drop tests or numerically. BAM has started a research project to get a better understanding about the behavior of prestressed bolt connection under loadings typical for drop tests.</abstract>
    <enrichment key="eventName">BAM-IRSN-Workshop 2021</enrichment>
    <enrichment key="eventPlace">Berlin, Germany</enrichment>
    <enrichment key="eventStart">08.09.2021</enrichment>
    <enrichment key="eventEnd">09.09.2021</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>Konrad Linnemann</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Transport packages</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Bolt connections</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Finite element analysis</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Experimental investigation</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>53261</id>
    <completedYear/>
    <publishedYear>2021</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">Polyurethane Foam in Impact Limiters - Experimental and Numerical Analysis</title>
    <abstract language="eng">Transport containers with radioactive material are usually shipped in Germany with wooden-filled impact limiters. Alternative energy-absorbing materials for this purpose can be polyurethane foam, for example. In order to perform adequate simulations with PU foam, experimental and numerical investigations must be carried out. With the help of a series of experimental test runs, a material model is developed, tested and compared in LS-Dyna.</abstract>
    <enrichment key="eventName">Technical Exchange IRSN – BAM: Transport &amp; Storage of Packages for Radioactive Material</enrichment>
    <enrichment key="eventPlace">Berlin, Germany</enrichment>
    <enrichment key="eventStart">08.09.2021</enrichment>
    <enrichment key="eventEnd">09.09.2021</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>Tobias Gleim</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Polyurethane Foam, Impact Limiters, PU-Foam</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>53262</id>
    <completedYear/>
    <publishedYear>2021</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">Polyurethane Foam for Impact Limiters - Theory and Experimental Data</title>
    <abstract language="eng">Transport containers with radioactive material are usually shipped in Germany with wooden-filled impact limiters. Alternative energy-absorbing materials for this purpose can be polyurethane foam, for example. In order to adequately determine the characteristics for a selected PU foam, various experimental investigations must be carried out. Classically, PU foam specimens are tested under uniaxial compression, confined compression and a tensile test. Using the experimental data and selected material models from the literature, the parameters can be determined to describe and apply material models.</abstract>
    <enrichment key="eventName">Technical Exchange Sandia – BAM: Transport &amp; Storage of Packages for Radioactive Material</enrichment>
    <enrichment key="eventPlace">Online meeting</enrichment>
    <enrichment key="eventStart">31.08.2021</enrichment>
    <enrichment key="eventEnd">31.08.2021</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>Tobias Gleim</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Polyurethane Foam</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Impact Limiters</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>PU-Foam</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>51962</id>
    <completedYear/>
    <publishedYear>2020</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>109</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>bookpart</type>
    <publisherName>IAEA Publishing Section</publisherName>
    <publisherPlace>Vienna, Austria</publisherPlace>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Methodology for a Safety Case of a Dual Purpose Cask for Storage and Transport of Spent Fuel</title>
    <abstract language="eng">Spent nuclear fuel which is generated in the operation of nuclear reactors needs to be safely managed following its removal from the reactor core. On-site power reactor storage pools were designed on the assumption that after a short period of time spent nuclear fuel would be removed for reprocessing and disposal or further storage elsewhere. The amount of highly radioactive spent fuel that needs to be stored over longer periods of time is growing and additional storage capacity is required. One of the widely used options for additional storage capacity is the use of casks for dry storage of spent fuel. Among various existing dry storage concepts, several Member States are utilizing a concept of dual purpose casks (DPCs). This publication provides practical advice on the structure and contents of a DPC integrated safety case with reference to existing IAEA requirements relevant to the licensing and use of transport and storage casks for spent fuel.</abstract>
    <parentTitle language="eng">IAEA TECDOC SERIES</parentTitle>
    <identifier type="isbn">978-92-0-131620-2</identifier>
    <identifier type="url">https://www.iaea.org/publications/10966/methodology-for-a-safety-case-of-a-dual-purpose-cask-for-storage-and-transport-of-spent-fuel</identifier>
    <identifier type="issn">1011–4289</identifier>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Bernhard Droste</author>
    <author>Dietmar Wolff</author>
    <author>A. Bevilacqua</author>
    <author>I. Reiche</author>
    <author>J. Harvey</author>
    <author>M. Hirose</author>
    <author>Y. Kumano</author>
    <author>P. McConnell</author>
    <author>T. Saegusa</author>
    <author>R. Einziger</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Dual Purpose Cask</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent Fuel</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Storage</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Transport</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="literaturgattung" number="">Verlagsliteratur</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>51988</id>
    <completedYear/>
    <publishedYear>2020</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">Embrittlement of Spent Fuel Claddings ‒ Results of Ring Compression Tests</title>
    <abstract language="eng">An established method for characterizing cladding material is the Ring Compression Test (RCT), where a small, cylindrical sample of the cladding tube is subjected to a compressive load. This test is a laboratory representation of a fuel rod load case and has shown a high susceptibility to failure under the occurrence of radial hydrides. Certain hydride morphologies and low temperature might even lead to brittle fracture at very small loads. The BAM research project BRUZL (Fracture mechanical analysis of spent fuel claddings during long-term dry interim storage) aims to study the results of quasi-static RCTs and to establish numerical models to gain an in-depth understanding of the stress state during such experiments. Using a fracture mechanical approach, cases of sudden failure during the RCT procedure are used to characterize the material behaviour and establish a failure criterion.&#13;
As an integral part of the project BRUZL, quasi-static Ring Compression Tests have been performed to identify all experimental details, which might be helpful for numerical modelling. Unirradiated samples of the cladding material ZIRLO® have been subjected to hydrogen charging and a thermo-mechanical treatment for radial hydride reorientation. Sample preparation, testing procedures, and analysis results are presented. A numerical model has been established and an elastic-plastic material model was derived from as-received RCT samples by inverse finite element analyses. Cohesive zone modelling has been implemented to reproduce sudden load drops during RCT.&#13;
The project is funded by the Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561.</abstract>
    <enrichment key="eventName">4th GRS Workshop on Safety of Extended Dry Storage of Spent Nuclear Fuel</enrichment>
    <enrichment key="eventPlace">Online meeting</enrichment>
    <enrichment key="eventStart">03.06.2020</enrichment>
    <enrichment key="eventEnd">04.06.2020</enrichment>
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    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <enrichment key="InvitedTalks">0</enrichment>
    <author>Kai Simbruner</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="">3.3 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>
  <doc>
    <id>52493</id>
    <completedYear/>
    <publishedYear>2021</publishedYear>
    <thesisYearAccepted/>
    <language>deu</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="deu">Aktuelle Entwicklungen und Prüfungen der BAM</title>
    <abstract language="deu">Es werden im Rahmen des Vortrages die aktuellen Prüftätigkeiten und Arbeitsschwerpunkte der BAM vorgestellt. Des Weiteren wird darüber berichtet was es Neues in der aktuellen Normen- und Richtlinienentwicklung gibt.</abstract>
    <enrichment key="eventName">Informationsveranstaltung zum Transport radioaktiver Stoffe</enrichment>
    <enrichment key="eventPlace">Online meeting</enrichment>
    <enrichment key="eventStart">14.04.2021</enrichment>
    <enrichment key="eventEnd">14.04.2021</enrichment>
    <enrichment key="opus.source">publish</enrichment>
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    <enrichment key="InvitedTalks">0</enrichment>
    <author>Frank Wille</author>
    <author>Steffen Komann</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Gefahrgutregeln der BAM</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Fallprüfungen</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Radioaktive Materialien</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Zulassung</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>
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    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
    <collection role="institutes" number="">3.0 Abteilungsleitung und andere</collection>
  </doc>
  <doc>
    <id>54742</id>
    <completedYear/>
    <publishedYear>2021</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>6</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName>American Society of Mechanical Engineers (ASME)</publisherName>
    <publisherPlace>New York, NY, USA</publisherPlace>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>1</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Investigation of the internal impact during a 9 m drop test of an accident-safe waste package</title>
    <abstract language="eng">The safety assessment of packages for the transport of radioactive material follows the IAEA regulations and guidance. The specified regulatory tests cover severe accidents and demonstrate the package containment system integrity.&#13;
Special attention must be drawn to the behaviour of the content which could move inside the package due to unpreventable gaps caused by the loading procedure and the structure of the content. A possible internal impact of the content which occurs during the drop tests onto the lid system is investigated. The IAEA regulations SSR-6 and the Guidance SSG-26, revised recently, consider input from Germany and France related to the significance of internal gaps.&#13;
In the context of a waste package design assessment, a model was equipped with a representative content to conduct a drop test with an internal impact. The weight and kinetic impact of this content covered all possible real contents. The objective of the test was to maximize the load onto the lid system and to prove the mechanical integrity by complying with the required leak tightness. The test was conducted&#13;
conservatively at a package temperature lower than -40 °C at the BAM Test Site Technical Safety.&#13;
This paper gives an overview of efforts to address internal gaps and their consequences, and the BAM efforts with the implementation of this topic into IAEA regulations and guidance material. The paper then focuses on the conduction of a drop test and investigation of internal component impact.</abstract>
    <parentTitle language="eng">Proceedings of the ASME 2021 Pressure Vessels &amp; Piping Conference (PVP2021)</parentTitle>
    <identifier type="isbn">978-0-7918-8535-2</identifier>
    <identifier type="doi">10.1115/PVP2021-60996</identifier>
    <enrichment key="eventName">ASME 2021 Pressure Vessels &amp; Piping Conference (PVP2021)</enrichment>
    <enrichment key="eventPlace">Online meeting</enrichment>
    <enrichment key="eventStart">13.07.2021</enrichment>
    <enrichment key="eventEnd">15.07.2021</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="date_peer_review">05.12.2022</enrichment>
    <author>Marko Nehrig</author>
    <author>Robert Scheidemann</author>
    <author>Frank Wille</author>
    <author>Viktor Ballheimer</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Internal gaps</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Drop test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>IAEA</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>54744</id>
    <completedYear/>
    <publishedYear>2021</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">Investigation of the internal impact during a 9 m drop test of an accident-safe waste package</title>
    <abstract language="eng">The safety assessment of packages for the transport of radioactive material follows the IAEA regulations and guidance. The specified regulatory tests cover severe accidents and demonstrate the package containment system integrity. Special attention must be drawn to the behaviour of the content which could move inside the package due to unpreventable gaps caused by the loading procedure and the structure of the content. A possible internal impact of the content which occurs during the drop tests onto the lid system is investigated. The IAEA regulations SSR-6 and the Guidance SSG-26, revised recently, consider input from Germany and France related to the significance of internal gaps.&#13;
In the context of a waste package design assessment, a model was equipped with a representative content to conduct a drop test with an internal impact. The weight and kinetic impact of this content covered all possible real contents. The objective of the test was to maximize the load onto the lid system and to prove the mechanical integrity by complying with the required leak tightness. The test was conducted conservatively at a package temperature lower than -40 °C at the BAM Test Site Technical Safety.&#13;
This paper gives an overview of efforts to address internal gaps and their consequences, and the BAM efforts with the implementation of this topic into IAEA regulations and guidance material. The paper then focuses on the conduction of a drop test and investigation of internal component impact.</abstract>
    <enrichment key="eventName">ASME 2021 Pressure Vessels &amp; Piping Conference (PVP2021)</enrichment>
    <enrichment key="eventPlace">Online meeting</enrichment>
    <enrichment key="eventStart">13.07.2021</enrichment>
    <enrichment key="eventEnd">15.07.2021</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="InvitedTalks">0</enrichment>
    <author>Marko Nehrig</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Drop test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Internal gaps</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>IAEA</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>
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    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>54856</id>
    <completedYear/>
    <publishedYear>2022</publishedYear>
    <thesisYearAccepted/>
    <language>deu</language>
    <pageFirst>112</pageFirst>
    <pageLast>114</pageLast>
    <pageNumber/>
    <edition/>
    <issue>3</issue>
    <volume>24</volume>
    <type>article</type>
    <publisherName>Bunsen-Gesellschaft</publisherName>
    <publisherPlace>Frankfurt</publisherPlace>
    <creatingCorporation>Vorstand der Deutschen Bunsen-Gesellschaft</creatingCorporation>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="deu">CASTOR in Warteposition, aber bitte sicher.</title>
    <abstract language="deu">CASTOR-Behälter sollen den Auswirkungen schwerster Unfälle standhalten und müssen dabei ihre Sicherheitsfunktionen beibehalten. Dazu zählen der sichere Einschluss des radioaktiven Inventars mit nachgewiesener Behälterintegrität und -dichtheit, die ausreichende Abschirmung der radioaktiven Strahlung, die Ableitung der von den Brennelementen ausgehenden Wärme und der Ausschluss des Entstehens einer nuklearen Kettenreaktion, d.h. die Kritikalitätssicherheit sind zu gewährleisten.&#13;
Zum Nachweis dieser Fähigkeiten werden die Behälter gegen Stoß, Aufprall, Durchstoßen, Feuer sowie beim Eintauchen in Wasser geprüft. Durch diese Tests werden Beanspruchungen aus potentiellen schweren Unfällen abdeckend simuliert. Die hierfür erforderlichen experimentellen Tests, aber auch die Prüfung und Entwicklung komplexer rechnerischer Simulationen des Behälterverhaltens unter Anwendung der Finite-Elemente-Methode erfolgen bei der Bundesanstalt für Materialforschung und -prüfung (BAM).&#13;
Die Basis bildet das Regelwerk der Internationalen Atom- und Energieorganisation (IAEO) zum sicheren Transport von radioaktiven Stoffen, welches international harmonisierte Schutzziele und Prüfanforderungen definiert. Die seit 60 Jahren praktizierte stetige Weiterentwicklung der Sicherheitsanforderungen trägt aktuellen Erkenntnissen, Erfahrungen und Risikoanalysen Rechnung. Die Brennelementbehälter werden mechanischen Prüfungen sowie Brand- und Wasserdruckprüfungen unterzogen. Hierdurch wird nachgewiesen, dass sie den Auswirkungen eines schweren Unfalls standhalten können.&#13;
Nur Behälter die zweifelsfrei die kumulativen mechanischen und thermischen Tests und auch die Wasserdruckprüfung mit Erfüllen aller Schutzziele bestehen, bekommen eine Zulassung, die es erlaubt die Behälter zu transportieren.</abstract>
    <parentTitle language="deu">Bunsen-Magazin, Zeitschrift der Deutschen Bunsen-Gesellschaft für physikalische Chemie</parentTitle>
    <identifier type="issn">1611-9479</identifier>
    <identifier type="doi">10.26125/73hj-nz53</identifier>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Frank Wille</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Kerntechnik</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Radioaktive Stoffe</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Transport</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Zwischenlagerung</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>IAEA</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Fallprüfung</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="literaturgattung" number="">Verlagsliteratur</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>55659</id>
    <completedYear/>
    <publishedYear>2021</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst/>
    <pageLast/>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>poster</type>
    <publisherName/>
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    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Temperature and Heat Flux Measurements in Fire Testing</title>
    <abstract language="eng">Packages for the transport of Spent Nuclear Fuel and high active radioactive waste are designed to withstand severe accidents. Specific mechanical and thermal tests are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials in order to cover these hypothetical severe accidents. The thermal test mainly consists of a 30 minute fully engulfing fire. Components such as the package impact limiters can lead to supplementary energy release during the thermal test as they might continue burning after the fire phase. Local heat flux into the package can occur. Measurement of heat flux into the package is of importance to evaluate component temperatures and review their acceptance. The usability of heat flux sensors and temperature measurement equipment has to be tested to apply them in impact limiter fire testing. Further questions arise such as :“Is the infrared camera a useful tool to determine heat flux at the boundary surface?”</abstract>
    <enrichment key="eventName">Technical Exchange IRSN – BAM Transport &amp; Storage of Packages for Radioactive Material</enrichment>
    <enrichment key="eventPlace">Berlin, Germany</enrichment>
    <enrichment key="eventStart">08.09.2021</enrichment>
    <enrichment key="eventEnd">09.09.2021</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Martin Feldkamp</author>
    <author>Thomas Quercetti</author>
    <author>Tobias Gleim</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Heat flux sensor</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>
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    <collection role="literaturgattung" number="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>55426</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">Experimental and numerical investigation of prestressed bolt connections under lateral displacements</title>
    <abstract language="eng">The containment system of transport packages for spent nuclear fuel and high-level waste usually includes bolted lids with metal gaskets. The packages are assessed to specific transport conditions which are specified in the IAEA safety standards SSR-6 (IAEA 2018). These transport conditions, especially the so-called accident conditions of transport, imply high dynamic loading on the lids and the bolt connections of the package. The response of the lid systems on the mechanical accident conditions is generally investigated by experimental drop tests or numerically, e.g., by finite element analyses. The interpretation of the drop test results for the verification of the numerical models is often not obvious due to the complex superposition of different effects in the real tests. BAM has started a research project to get a better understanding about the behavior of prestressed bolt connections under loadings typical for these drop tests. In this context an experimental test set-up was developed to investigate the response of a single bolt connection under a prescribed lateral displacement of clamped parts. The bolt is instrumented by strain gauges to get the pretensional, the torsional and the bending stress in the bolt shank. Furthermore, the lateral movement and the tilt of the bolt head is measured during the test. A finite element model of the test set-up has been created in Abaqus FEA (Simulia 2021). The very detailed instrumentation of the test set-up shall give the opportunity to investigate and validate the numerical model. The aim of this paper is to give an overview about the proposed research project and to present first results.</abstract>
    <enrichment key="eventName">SMIRT26 - 26th International Conference on Structural Mechanics in Reactor Technology</enrichment>
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    <author>Konrad Linnemann</author>
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      <language>eng</language>
      <type>uncontrolled</type>
      <value>Transport packages</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Bolt connections</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Finite element analysis</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Experimental investigation</value>
    </subject>
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    <title language="eng">New Sco-III regulations to ship large objects as surface contaminated objects</title>
    <abstract language="eng">The decommissioning or refurbishment of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, and coolant pumps, to list the major contributors. These components or objects are large in size and mass, measuring up to approximately 6 meters in diameter, up to 20 meters in length, and weighing over 400 000 kg. In many situations, the components are transported off-site to a storage, disposal or recycling/treatment facility. Previously, many large objects had to be transported under special arrangement.&#13;
The latest 2018 edition of the International Atomic Energy Agency (IAEA) Regulations for the Safe Transport of Radioactive Material, No. SSR-6 [1], incorporates regulations for the shipment of large objects as a new category of surface contaminated object, SCO-III, based on the IAEA “performance package” concept. This paper provides background and practical guidance on these regulations.&#13;
Additionally, the experiences of BAM with the appoval of two steam converters of the NPP Lingen are presented as the first approval process for SCO-III objects in Germany.&#13;
The primary additions to SSR-6 include SCO-III classification and requirements, approval and administrative requirements for the new classification, and the addition of SCO-III to the proper shipping&#13;
name for UN 2913. Advisory material drafted for the new requirements will be included in the next revision of SSG-26, Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, expected to be published soon.&#13;
Note that at this time the proposed provisions for large objects do not include components such as reactor vessels, due to the more limited experience and greater radioactivity levels. The SCO-III concept lays the groundwork and may be extended to cover other large objects that are classified as low specific activity (LSA) material in the future.</abstract>
    <parentTitle language="eng">Proceedings der Kerntechnik 2022</parentTitle>
    <enrichment key="eventName">Kerntechnik 2022</enrichment>
    <enrichment key="eventPlace">Leipzig, Germany</enrichment>
    <enrichment key="eventStart">21.06.2022</enrichment>
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    <author>Konrad Linnemann</author>
    <author>Steffen Komann</author>
    <author>Frank Wille</author>
    <author>I. Reiche</author>
    <author>J. Ramsay</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Transport packages</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Radioactive materials</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>SCO-III</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Large objects</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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    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
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    <title language="deu">Neue SCO-III Regularien für den Transport grosser Gegenstände als oberflächenkontaminierte Objekte</title>
    <abstract language="eng">The decommissioning or refurbishment of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, and coolant pumps, to list the major contributors. These components or objects are large in size and mass, measuring up to approximately 6 meters in diameter, up to 20 meters in length, and weighing over 400 000 kg. In many situations, the components are transported off-site to a storage, disposal or recycling/treatment facility. Previously, many large objects had to be transported under special arrangement.&#13;
The latest 2018 edition of the International Atomic Energy Agency (IAEA) Regulations for the Safe Transport of Radioactive Material, No. SSR-6 [1], incorporates regulations for the shipment of large objects as a new category of surface contaminated object, SCO-III, based on the IAEA “performance package” concept. This paper provides background and practical guidance on these regulations. Additionally, the experiences of BAM with the appoval of two steam converters of the NPP Lingen are presented as the first approval process for SCO-III objects in Germany.&#13;
The primary additions to SSR-6 include SCO-III classification and requirements, approval and administrative requirements for the new classification, and the addition of SCO-III to the proper shipping name for UN 2913. Advisory material drafted for the new requirements will be included in the next revision of SSG-26, Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, expected to be published soon. Note that at this time the proposed provisions for large objects do not include components such as reactor vessels, due to the more limited experience and greater radioactivity levels. The SCO-III concept lays the groundwork and may be extended to cover other large objects that are classified as low specific activity (LSA) material in the future.</abstract>
    <enrichment key="eventName">Kerntechnik 2022</enrichment>
    <enrichment key="eventPlace">Leipzig, Germany</enrichment>
    <enrichment key="eventStart">21.06.2022</enrichment>
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    <author>Konrad Linnemann</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Transportbehälter</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Radioaktive Stoffe</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Oberflächenkontaminiert</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>SCO-III</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
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  </doc>
  <doc>
    <id>58088</id>
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    <publishedYear>2023</publishedYear>
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    <language>eng</language>
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    <title language="eng">Test facilities for transport and storage packages at BAM</title>
    <abstract language="eng">BAM acts as authority and for service in safety assessment of packages for transport and storage of radioactive materials. We offer extensive test capabilities and application of analytical methods for design verification and simulation for all types of packages for the transport and storage of radioactive materials according with the international IAEA Regulations for the safe transport and for national storage acceptance criteria. BAM operates several test facilities for drop and stacking testing, leak testing and thermal testing. The large drop test tower allows dropping full-scale specimens up to 200,000 kg in any drop orientation as requested. The comprehensive test facilities combined with long-term experience, newest equipment and measurement devices according to the latest state-of-the-art technology ensures realisation of complex test campaigns for package safety evaluation.&#13;
Beyond that, non-destructive and destructive material test devices and experts are available. Equipment and application of all kinds of typical measurement categories can be offered for testing campaigns.&#13;
In recent years we performed testing of full-scale type B package models with complex handling and preparation procedures. The results were contributed for different package design approval procedures. Type A packages mainly designed for medical related transport purposes, were continuously tested according to the transport regulations over recent years as well. Moreover, we work on research topics with relevance to package safety. The mechanical behaviour of lid closure systems under transport and storage conditions and the thermal behaviour of impact limiters were recently of special importance for the assessment competencies of BAM and were investigated under use of our test facilities.&#13;
The paper describes the test facilities and capabilities for package design safety evaluation at BAM and shows examples from our recent work.</abstract>
    <enrichment key="eventName">PATRAM 2022, 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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    <author>Frank Wille</author>
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      <language>eng</language>
      <type>uncontrolled</type>
      <value>Transport</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Drop testing</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Fire test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>IAEA</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Package</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
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  </doc>
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    <thesisYearAccepted/>
    <language>eng</language>
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    <title language="eng">New SCO-iii regulations to ship large comonents as surface contaminated objects</title>
    <abstract language="eng">The decommissioning or refurbishment of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, and coolant pumps, to list the major contributors. These components or objects are large in size and mass, measuring up to approximately 6 meters in diameter, up to 20 meters in length, and weighing over 400 000 kg. In many situations, the components are transported off-site to a storage, disposal or recycling/treatment facility. Previously, many large objects had to be transported under special arrangement.</abstract>
    <parentTitle language="eng">PATRAM22</parentTitle>
    <enrichment key="eventName">PATRAM22</enrichment>
    <enrichment key="eventPlace">Juan les Pins, France</enrichment>
    <enrichment key="eventStart">11.06.2023</enrichment>
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    <author>Steffen Komann</author>
    <author>Konrad Linnemann</author>
    <author>Frank Wille</author>
    <author>Ingo Reiche</author>
    <author>J. Ramsay</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Gefahrgut</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Radioaktive Stoffe</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Rückbau kerntechnischer Anlagen</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Transport</value>
    </subject>
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  </doc>
  <doc>
    <id>57781</id>
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    <publishedYear>2023</publishedYear>
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    <title language="eng">Ageing aspect in design approval of special form radioactive material</title>
    <abstract language="eng">In accordance with the IAEA transport regulations, the design of special form radioactive material (SFRM) shall resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval includes besides the program for physical tests (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance, and inspection.&#13;
SFRM source design plus management system shall ensure, that every specimen of the approved design is able to survive the severe mechanical and thermal tests at any time of its SFRM-working life.&#13;
Due to the long-term use of SFRM designs in most cases, the assessment of the source ageing behavior is an important aspect in the approval procedure. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. Besides of radioactive content, corrosion is a main factor for possible SFRM design degradation.&#13;
Although the IAEA Advisory Material SSG-26 already implies an indication of the need for considering ageing mechanisms, suitable amendments in the regulatory requirements of SSR-6 should be introduced to make the approval procedure more transparent and help to reduce rounds of questions by the authority. A supplementary requirement for considering of ageing mechanisms could be a helpful contribution to an international harmonization of the approval procedure.&#13;
This paper will describe major influencing factors to be considered to assess the ageing behavior of a SFRM design and will identify the need for a regulatory specification of a SFRM-working life as basis for the assessment of the SFRM design regarding time-dependent weakening. A proposal for an explicit requirement for consideration of ageing mechanisms in safety assessment of SFRM, which should be considered in the ongoing SSR-6 revision cycle, will be explained.</abstract>
    <enrichment key="eventName">PATRAM22 Conference</enrichment>
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    <enrichment key="eventStart">11.06.2023</enrichment>
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    <author>Annette Rolle</author>
    <author>Steffen Komann</author>
    <author>Frank Wille</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Radioaktive Stoffe</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Beförderung</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Zulassung</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Strahlenschutz</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
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  </doc>
  <doc>
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    <publishedYear>2023</publishedYear>
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    <title language="eng">Ageing aspect in design approval of special form radioactive material</title>
    <abstract language="eng">In accordance with the IAEA transport regulations, the design of special form radioactive material (SFRM) shall resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval includes besides the program for physical tests (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance, and inspection.&#13;
SFRM source design plus management system shall ensure, that every specimen of the approved design is able to survive the severe mechanical and thermal tests at any time of its SFRM-working life.&#13;
Due to the long-term use of SFRM designs in most cases, the assessment of the source ageing behavior is an important aspect in the approval procedure. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. Besides of radioactive content, corrosion is a main factor for possible SFRM design degradation.&#13;
Although the IAEA Advisory Material SSG-26 already implies an indication of the need for considering ageing mechanisms, suitable amendments in the regulatory requirements of SSR-6 should be introduced to make the approval procedure more transparent and help to reduce rounds of questions by the authority. A supplementary requirement for considering of ageing mechanisms could be a helpful contribution to an international harmonization of the approval procedure.&#13;
This paper will describe major influencing factors to be considered to assess the ageing behavior of a SFRM design and will identify the need for a regulatory specification of a SFRM-working life as basis for the assessment of the SFRM design regarding time-dependent weakening. A proposal for an explicit requirement for consideration of ageing mechanisms in safety assessment of SFRM, which should be considered in the ongoing SSR-6 revision cycle, will be explained.</abstract>
    <parentTitle language="eng">Proceedings of the 20th International Symposium on the Packaging and Transportation of Radioactive Materials iio</parentTitle>
    <enrichment key="eventName">PATRAM22 Conference</enrichment>
    <enrichment key="eventPlace">Juan-Les-Pins, Antibes, France</enrichment>
    <enrichment key="eventStart">11.06.2023</enrichment>
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    <author>Annette Rolle</author>
    <author>Steffen Komann</author>
    <author>Frank Wille</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ttransport</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Radioactive material</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Sealed sources</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ageing</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
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    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
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    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="deu">New SCO-III regulations to ship large components as surface contaminated objects</title>
    <abstract language="deu">The decommissioning or refurbishment of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, and coolant pumps, to list the major contributors. These components or objects are large in size and mass, measuring up to approximately 6 meters in diameter, up to 20 meters in length, and weighing over 400 000 kg. In many situations, the components are transported off-site to a storage, disposal or recycling/treatment facility. Previously, many large objects had to be transported under special arrangement</abstract>
    <enrichment key="eventName">PATRAM22 Konferenz</enrichment>
    <enrichment key="eventPlace">Juan-les-Pins, France</enrichment>
    <enrichment key="eventStart">11.06.2023</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>Steffen Komann</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Radioaktive Stoffe</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Rückbau kerntechnischer Anlagen</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Transport</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>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>
    <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>57967</id>
    <completedYear/>
    <publishedYear>2023</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>12</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation>INMM</creatingCorporation>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Test Facilities for Radioactive Materials Transport and Storage Packagings at BAM</title>
    <abstract language="eng">BAM acts as authority and for service in safety assessment of packages for transport and storage of radioactive materials. We offer extensive test capabilities and application of analytical methods for design verification and simulation for all types of packages for the transport and storage of radioactive materials according with the international IAEA Regulations for the safe transport and for national storage acceptance criteria. BAM operates several test facilities for drop and stacking testing, leak testing and thermal testing. The large drop test tower allows dropping full-scale specimens up to 200,000 kg in any drop orientation as requested. The comprehensive test facilities combined with long-term experience, newest equipment and measurement devices according to the latest state-of-the-art technology ensures realisation of complex test campaigns for package safety evaluation.&#13;
Beyond that, non-destructive and destructive material test devices and experts are available. Equipment and application of all kinds of typical measurement categories can be offered for testing campaigns.&#13;
In recent years we performed testing of full-scale type B package models with complex handling and preparation procedures. The results were contributed for different package design approval procedures. Type A packages mainly designed for medical related transport purposes, were continuously tested according to the transport regulations over recent years as well. Moreover, we work on research topics with relevance to package safety. The mechanical behaviour of lid closure systems under transport and storage conditions and the thermal behaviour of impact limiters were recently of special importance for the assessment competencies of BAM and were investigated under use of our test facilities.&#13;
The paper describes the test facilities and capabilities for package design safety evaluation at BAM and shows examples from our recent work.</abstract>
    <parentTitle language="eng">PATRAM 2022, Proceedings of the 20th International Symposium on the Packaging and Transportation of Radioactive Materials</parentTitle>
    <enrichment key="eventName">PATRAM 2022, 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>
    <enrichment key="eventEnd">15.06.2023</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Frank Wille</author>
    <author>Thomas Quercetti</author>
    <author>Robert Scheidemann</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>IAEA</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Fire test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Drop testing</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Transport</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Package</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>57719</id>
    <completedYear/>
    <publishedYear>2023</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">Fire Influence to Wood Filled Impact Limiters - Implications for the Package Design Safety Case</title>
    <abstract language="eng">Impact limiters with wooden components are widely used in the design of packages for transportation of radioactive material. In most designs, the wood is encapsulated with steel sheets. Impact limiters mainly determine the mechanical and thermal behaviour of the package in accident conditions of transport in accordance with the IAEA Regulations.  &#13;
In context with research and development for package design approval competence, the thermal behaviour of heavy-weight packages was investigated at BAM with an artificially pre-damaged generic impact limiter design. Within this first investigation, the pre-damaged impact limiter with a diameter of 2.3 meters was mounted on a water tank simulating the thermal capacity of a cask during the fire test. The water tank is part of a water circulating system built of several components such as pump, heater, cooler, sliding valve, flow meter, thermocouples and control unit in order to measure the heat flux.&#13;
Furthermore, the investigations focus on the effects this additional heat generation would have on the cask and especially on the lid-closure system with the gasket. &#13;
The results of these experiments could find consideration in the safety case of the transport packages of radioactive material using wood filled impact limiters.</abstract>
    <enrichment key="eventName">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>
    <enrichment key="eventEnd">15.06.2023</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Andre Musolff</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Fire test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Impact limiter</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Mechanical and thermal behaviour</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>59421</id>
    <completedYear/>
    <publishedYear>2023</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>11</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Welding Seam Safety Evaluation in a Thick-Walled Type B Transport Package</title>
    <abstract language="eng">The safety demonstration of heavy weight type B transport packages used for storing spent nuclear fuel (SNF) or radioactive waste is ensured by a combination of physical testing and numerical calculations. While experiments are performed in accordance with the IAEA regulations for selected drop scenarios, Finite-Element-Method (FEM) simulations are used to predict the most damaging case and to investigate additional drop positions. BAM as competent authority in Germany has performed different investigations of a welding seam for a typical large transport package made of A508 forged steel, where the bottom plate is welded to the cylindrical shell. The package has a mass of approx. 120 t. Results of physical drop tests with a full-scale model and accompanying preliminary FEM simulations are presented to determine the decisive stresses in the welding seam. A drop test only represents one set of a package and test parameters. A further parameter analysis is considered to account for allowable variations of packaging properties (e.g. resulting from the manufacturing process) and, based on IAEA requirements, the temperature dependence of the material behaviour. The results of the stress analyses from the drop test and the simulation form the basis and provide the input parameters for a fracture mechanics analysis. In addition to the IAEA specifications, further standards are taken into account for an in-depth investigation, see R6 [1], BS 7910 [2] and API 579-1/ASME FFS1 [3]. All the above-mentioned standards require a manufacturer-specific defect analysis with respect to size and position. Both result from the welding process and the following heat treatment regime. The maximum defect sizes are ensured with non-destructive test methods (such as ultrasonic or particle methods) as integral part of the manufacturing process of the welding seam. Another important parameter in the welding process is the residual stress (secondary stress). The combination of the primary and secondary stress determines the total stress in the welding seam. The most damaging case of the welding seam is determined and evaluated with help of the abovementioned standards and taking into account the IAEA requirements with respect to defect sizes, material properties, primary and residual stress, yield strength etc.</abstract>
    <parentTitle language="eng">Proceedings of the 20th International Symposium on the Packaging and Transportation of Radioactive Materials</parentTitle>
    <enrichment key="eventName">PATRAM 22 - The International Symposium on the Packaging and Transportation of Radioactive Materials</enrichment>
    <enrichment key="eventPlace">Juan-Les-Pins, Antibes</enrichment>
    <enrichment key="eventStart">11.06.2023</enrichment>
    <enrichment key="eventEnd">16.06.2023</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Tobias Gleim</author>
    <author>Martin Neumann</author>
    <author>Konrad Linnemann</author>
    <author>Frank Wille</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Welding</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Transport Package</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Fracture Mechanics</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>59679</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>10</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName>IASMiRT</publisherName>
    <publisherPlace/>
    <creatingCorporation>International Association for Structural Mechanics in Reactor Technology (IASMiRT)</creatingCorporation>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Evaluation of Heat Fluxes in Fire Reference Test Conducted in BAM Propane Gas Fire Test Facility</title>
    <abstract language="eng">Packages for the transport of intermediate- and high-level radioactive waste are designed to withstand severe accidents. The International Atomic Energy Agency (IAEA) has established specific mechanical and thermal tests. Packages for the transport of radioactive material must withstand these tests to comply with the Regulations for the Safe Transport of Radioactive Materials IAEA [IAEA (2018)]. A fire reference package was developed with the primary objective to demonstrate that the fire meets the regulatory requirements. Another aim is to characterise the boundary conditions of the actual fire as input parameters for thermo-mechanical simulations. A simple method to characterise the boundary conditions of a real steady state fire with a fire reference package is presented. The thermal test mainly consists of a 30 minute fully engulfing 800°C pool fire or an equally severe fire, such as a propane gas fire. The fire reference tests are performed prior to the actual fire test with the real package. The heat fluxes into the package depend substantially on the fire characteristics and the surface temperature of the package. To investigate local and overall heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for repeated use. The fire reference package presented in this paper represents the outer geometry of a small transport container for radioactive material and is used as a device in civil engineering. It is designed as a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 182 mm and a diameter of 102 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material. Open-air fire tests were performed in a BAM propane gas fire test facility with the fire reference package. The measured temperatures are used to determine the changes of heat fluxes into the fire reference package in relation to the package surface temperature. The calculated heat fluxes allow its fitting to express the thermal exposure as simple mathematical boundary condition. Therefore, in a first approach, fire properties such as adiabatic surface temperature (AST) as proposed by Wickström et al. (2007), convection coefficient and emissivity are determined mathematically fitting the heat flux development presented in this paper. The evaluated results provide an initial picture of local fire characteristics of the conducted propane gas fire and are a further development of previous works from Feldkamp et al. (2020). The results can be used in thermal and thermo-mechanical models to simulate the load on the real transport package in fire. The test shows that the examined propane gas fire covers the IAEA-fire over a wide range of surface temperatures with the chosen test stand parameters.</abstract>
    <parentTitle language="eng">Transactions of SMiRT 27 (27th conference on Structural Mechanics in Reactor Technology)</parentTitle>
    <enrichment key="eventName">SMiRT 27 (27th conference on Structural Mechanics in Reactor Technology)</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>Martin Feldkamp</author>
    <author>Thomas Quercetti</author>
    <author>Tobias Gleim</author>
    <author>Frank Wille</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Fire</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Propane</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Heat Flux</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Fire Reference</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>59680</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">Evaluation of Heat Fluxes in Fire Reference Test Conducted in BAM Propane Gas Fire Test Facility</title>
    <abstract language="eng">Packages for the transport of intermediate- and high-level radioactive waste are designed to withstand severe accidents. The International Atomic Energy Agency (IAEA) has established specific mechanical and thermal tests. Packages for the transport of radioactive material must withstand these tests to comply with the Regulations for the Safe Transport of Radioactive Materials IAEA [IAEA (2018)]. A fire reference package was developed with the primary objective to demonstrate that the fire meets the regulatory requirements. Another aim is to characterise the boundary conditions of the actual fire as input parameters for thermo-mechanical simulations. A simple method to characterise the boundary conditions of a real steady state fire with a fire reference package is presented. The thermal test mainly consists of a 30 minute fully engulfing 800°C pool fire or an equally severe fire, such as a propane gas fire. The fire reference tests are performed prior to the actual fire test with the real package. The heat fluxes into the package depend substantially on the fire characteristics and the surface temperature of the package. To investigate local and overall heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for repeated use. The fire reference package presented in this paper represents the outer geometry of a small transport container for radioactive material and is used as a device in civil engineering. It is designed as a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 182 mm and a diameter of 102 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material. Open-air fire tests were performed in a BAM propane gas fire test facility with the fire reference package. The measured temperatures are used to determine the changes of heat fluxes into the fire reference package in relation to the package surface temperature. The calculated heat fluxes allow its fitting to express the thermal exposure as simple mathematical boundary condition. Therefore, in a first approach, fire properties such as adiabatic surface temperature (AST) as proposed by Wickström et al. (2007), convection coefficient and emissivity are determined mathematically fitting the heat flux development presented in this paper. The evaluated results provide an initial picture of local fire characteristics of the conducted propane gas fire and are a further development of previous works from Feldkamp et al. (2020). The results can be used in thermal and thermo-mechanical models to simulate the load on the real transport package in fire. The test shows that the examined propane gas fire covers the IAEA-fire over a wide range of surface temperatures with the chosen test stand parameters.</abstract>
    <enrichment key="eventName">SMiRT 27 (27th conference on Structural Mechanics in Reactor Technology)</enrichment>
    <enrichment key="eventPlace">Yokohama, Japan</enrichment>
    <enrichment key="eventStart">03.03.2024</enrichment>
    <enrichment key="eventEnd">08.03.2024</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>Martin Feldkamp</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Fire</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Propane</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Fire Reference Test</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>59378</id>
    <completedYear/>
    <publishedYear>2022</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>7</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName>IASMiRT</publisherName>
    <publisherPlace/>
    <creatingCorporation>International Association for Structural Mechanics in Reactor Technology (IASMiRT)</creatingCorporation>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Experimental and numerical investigation of prestressed bolt connections under lateral displacements</title>
    <abstract language="eng">The containment system of transport packages for spent nuclear fuel and high-level waste usually includes bolted lids with metal gaskets. The packages are assessed to specific transport conditions which are specified in the IAEA safety standards SSR-6 (IAEA 2018). These transport conditions, especially the so-called accident conditions of transport, imply high dynamic loading on the lids and the bolt connections of the package. The response of the lid systems on the mechanical accident conditions is generally investigated by experimental drop tests or numerically, e.g., by finite element analyses. The interpretation of the drop test results for the verification of the numerical models is often not obvious due to the complex superposition of different effects in the real tests. BAM has started a research project to get a better understanding about the behavior of prestressed bolt connections under loadings typical for these drop tests. In this context an experimental test set-up was developed to investigate the response of a single bolt connection under a prescribed lateral displacement of clamped parts. The bolt is instrumented by strain gauges to get the pretensional, the torsional and the bending stress in the bolt shank. Furthermore, the lateral movement and the tilt of the bolt head is measured during the test. A finite element model of the test set-up has been created in Abaqus FEA (Simulia 2021). The very detailed instrumentation of the test set-up shall give the opportunity to investigate and validate the numerical model. The aim of this paper is to give an overview about the proposed research project and to present first results.</abstract>
    <parentTitle language="eng">Transactions of SMIRT 26 (26th conference on Structural Mechanics in Reactor Technology)</parentTitle>
    <identifier type="url">https://www.lib.ncsu.edu/resolver/1840.20/40614</identifier>
    <enrichment key="eventName">SMIRT 26 (26th conference on Structural Mechanics in Reactor Technology)</enrichment>
    <enrichment key="eventPlace">Potsdam, Germany</enrichment>
    <enrichment key="eventStart">10.07.2022</enrichment>
    <enrichment key="eventEnd">15.07.2023</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Konrad Linnemann</author>
    <author>Thomas Quercetti</author>
    <author>Jens Sterthaus</author>
    <author>Viktor Ballheimer</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Bolt connections</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Finite element anaylsis</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Experimental testing</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Transport package for radioactive materials</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>58577</id>
    <completedYear/>
    <publishedYear>2023</publishedYear>
    <thesisYearAccepted/>
    <language>deu</language>
    <pageFirst>16</pageFirst>
    <pageLast>18</pageLast>
    <pageNumber/>
    <edition/>
    <issue>9</issue>
    <volume>2023</volume>
    <type>article</type>
    <publisherName>Heinrich Vogel</publisherName>
    <publisherPlace>München</publisherPlace>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="deu">Neue Regel für das Alterungsmanagement</title>
    <abstract language="deu">Die Regelwerksanforderungen für Verpackungen für den Transport radioaktiver Stoffe basieren auf den Empfehlungen der IAEA und werden über die verkehrsträgerspezifischen internationalen Regelwerke in nationales Recht umgesetzt.&#13;
Um die Anforderungen des Regelwerkes zu erläutern, entwickelt die Bundesanstalt für Materialforschung und -prüfung (BAM) Gefahrgutregeln (BAM-GGR), welche die Entwickler, Hersteller und Betreiber von Behältern über die von der BAM angewandten Verfahren bei der Umsetzung gesetzlicher Bestimmungen informieren und unterstützen sollen.</abstract>
    <parentTitle language="deu">Gefahrgut</parentTitle>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Steffen Komann</author>
    <author>Frank Wille</author>
    <author>Sven Schubert</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Gefahrgut</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Radioaktive Stoffe</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Alterung</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Transport</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="literaturgattung" number="">Verlagsliteratur</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>58436</id>
    <completedYear/>
    <publishedYear>2023</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">Investigations of Aged Metal Seals for Transport Package Safety Assessment</title>
    <abstract language="eng">Acceptable limits for activity release from transport casks for high-level radioactive material specified in the IAEA regulations must be kept by the integrity of cask body and the cask sealing system. BAM as the German competent authority for mechanical, thermal and containment assessment of packages liable for approval verifies the activity release compliance with the regulatory limits. One of the fundamental aspects in assessment is the specification of conservative package design leakage rates.&#13;
To ensure the required package tightness for both, storage, and transport of the cask before and after storage usually metal seals of the Helicoflex® Type are used. Due to the long-term use the seal behavior is influenced by temperature and time.  The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport specified in the regulations can have a significant effect on the leak tightness of the sealing system. Whereas the safety for application of new, non- aged Helicoflex® seals is verified sufficiently, there are still technical data gaps concerning the efficiency of aged Helicoflex® seals. &#13;
BAM performed experiments to learn more about the sealing efficiency of aged Helicoflex® seals with Aluminum and Silver outer jackets. The seals were compressed in test-flanges and for artificial ageing the complete flange systems were stored in an oven for several month at a high temperature. During the compression and decompression tests after the aging, load-deformation characteristics of the seals, and leakage rates were measured. With these tests a load situation was simulated, which can occur in the regulatory drop test of the cask: Under high impact loads the bolted lid can lift a little for a short moment, allowing a little movement of the seal, so that the contact area can change before compressing again. &#13;
The poster presentation will show details about test conditions and first results.</abstract>
    <enrichment key="eventName">Interdisciplinary research symposium on the safety of nuclear disposal practices safeND2023</enrichment>
    <enrichment key="eventPlace">Berlin, Germany</enrichment>
    <enrichment key="eventStart">13.09.2023</enrichment>
    <enrichment key="eventEnd">15.09.2023</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Annette Rolle</author>
    <author>Tino Neumeyer</author>
    <author>Viktor Ballheimer</author>
    <author>Frank Wille</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Radioactive material</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Sealing</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ageing</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Leaktightnes</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>57251</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">Fire Reference Tests for Qualification of IAEA Fire</title>
    <abstract language="eng">A small cylindrical fire reference package was designed. A fire test setup was created using the fire reference package in accordance with IAEA standards. Four propane gas fire tests were conducted to determine heat fluxes into the fire reference package. The initial test setup was modified to create the final design of the fire test facility. This allowed for the heat flux to be adjusted to meet the regulatory IAEA fire qualification criteria. Furthermore, a numerical model of the fire reference package was created using boundary conditions derived from the experimental data. The simulation results demonstrated good agreement with the experimental data and provided additional insights.</abstract>
    <enrichment key="eventName">Technical Exchange IRSN – BAM Transport &amp; Storage of Packages for Radioactive Material</enrichment>
    <enrichment key="eventPlace">Cadarache, France</enrichment>
    <enrichment key="eventStart">13.10.2022</enrichment>
    <enrichment key="eventEnd">14.10.2022</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>Martin Feldkamp</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Fire</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>IAEA</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Propane</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>60555</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">Development of a Hydrogen Fire Test Rig for Thermal Testing</title>
    <abstract language="eng">A new hydrogen-based test rig is being developed for a feasibility study on the use of hydrogen as an energy source for thermal testing in accordance with the IAEA SSR-6 guidelines for transport packages containing radioactive materials. &#13;
The test rig will be capable of combusting hydrogen for a wide range of different burner geometries, mass flows and if necessary hydrogen blends. As this type of fire test according to the IAEA boundary conditions (SSR-6 &amp; SSG-26) does not yet exist, a large number of preliminary investigations, safety assessments and calculations must be carried out in order to develop a viable concept for hydrogen fires.</abstract>
    <enrichment key="eventName">1st JF BMUV Förderprogramm</enrichment>
    <enrichment key="eventPlace">Berlin, Germany</enrichment>
    <enrichment key="eventStart">22.05.2024</enrichment>
    <enrichment key="eventEnd">22.05.2024</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Maximilian Naster</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>IAEA Fire</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>H2 Fire</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>IAEA Regulations</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>60505</id>
    <completedYear/>
    <publishedYear>2021</publishedYear>
    <thesisYearAccepted/>
    <language>deu</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="deu">Transport nach Zwischenlagerung, Besonderheiten des letzten Transports</title>
    <abstract language="deu">Transportierbarkeit von Transport- und Lagerbehältern für radioaktive Stoffe, welche nach eine längerfristigen trockenen Zwischenlagerung&#13;
transportiert werden müssen. Überblick über die aktuellen Regelwerke und Richtlinien sowie die Zeitpläne für das Finden eines Endlagers für hochradioaktive Stoffe und die damit verbundene verlängerte Zwischenlagerung. Des Weiteren werden das für die verlängerte Zwischenlagerung notwendige Alterungsmanagement und die noch erforderlichen Forschungstätigkeiten vorgestellt.</abstract>
    <enrichment key="eventName">BGZ Fachworkshop Zwischenlagerung</enrichment>
    <enrichment key="eventPlace">Berlin, Germany</enrichment>
    <enrichment key="eventStart">16.11.2021</enrichment>
    <enrichment key="eventEnd">17.11.2021</enrichment>
    <enrichment key="InvitedTalks">1</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Frank Wille</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Transportbehälter</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Radioaktive Stoffe</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Transport</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Zwischenlagerung</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>IAEA</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>60506</id>
    <completedYear/>
    <publishedYear>2023</publishedYear>
    <thesisYearAccepted/>
    <language>deu</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="deu">Verkehrsrechtliche Betrachtungen Transport- und Lagerbehälter - Wo geht die Reise hin?</title>
    <abstract language="deu">Transportierbarkeit von Transport- und Lagerbehältern für radioaktive Stoffe, welche nach eine längerfristigen trockenen Zwischenlagerung&#13;
transportiert werden müssen. Überblick über die aktuellen Regelwerke und Richtlinien sowie die Zeitpläne für das Finden eines Endlagers für hochradioaktive Stoffe und die damit verbundene verlängerte Zwischenlagerung. Des Weiteren werden das für die verlängerte Zwischenlagerung notwendige Alterungsmanagement und die noch erforderlichen Forschungstätigkeiten vorgestellt.</abstract>
    <enrichment key="eventName">10. Symposium Lagerung und Transport radioaktiver Stoffe</enrichment>
    <enrichment key="eventPlace">Hannover, Germany</enrichment>
    <enrichment key="eventStart">05.09.2023</enrichment>
    <enrichment key="eventEnd">06.09.2023</enrichment>
    <enrichment key="InvitedTalks">1</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Frank Wille</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Transportbehälter</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Radioaktive Stoffe</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Transport</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Zwischenlagerung</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>IAEA</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>60554</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">Development of a Hydrogen Fire Test Rig for Thermal Testing</title>
    <abstract language="eng">A new hydrogen-based test rig is being developed for a feasibility study on the use of hydrogen as an energy source for thermal testing in accordance with the IAEA SSR-6 guidelines for transport packages containing radioactive materials. &#13;
The test rig will be capable of combusting hydrogen for a wide range of different burner geometries, mass flows and if necessary hydrogen blends. As this type of fire test according to the IAEA boundary conditions (SSR-6 &amp; SSG-26) does not yet exist, a large number of preliminary investigations, safety assessments and calculations must be carried out in order to develop a viable concept for hydrogen fires.</abstract>
    <enrichment key="eventName">Technical Exchange IRSN – BAM Transport &amp; Storage of Packages for Radioactive Material</enrichment>
    <enrichment key="eventPlace">Berlin, Germany</enrichment>
    <enrichment key="eventStart">04.06.2024</enrichment>
    <enrichment key="eventEnd">05.06.2024</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Maximilian Naster</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>IAEA Fire</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>H2 Fire</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>IAEA Regulations</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>60847</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">Experimental And Numerical Analyses For The Evaluation Of Heat Fluxes Of A Fire Reference Test</title>
    <abstract language="eng">Packages for the transport of high-level radioactive material are designed to endure severe accidents. To obtain approval, these transport packages must adhere to the specification-based criteria of the IAEA SSR-6 [1]. To ensure compliance with these requirements, specific mechanical and thermal tests need to be appointed with respect to the package type. Typically, IAEA SSR-6 [1] prescribes mechanical tests followed by a thermal fire test as part of a cumulative test sequence. To approve the fire test, BAM uses a reference package representing the original package in its outer geometry for characterizing the actual fire and its impact on the package. This serves a dual purpose: it allows the precise adjustment of experimental parameters for the package design to be approved, and it provides input parameters for thermomechanical simulations, cf. [2]. With the help of this methodology for characterizing boundary conditions of the package, temperature evolutions within the reference package can be studied in a Finite-Element Analyses (FEA). This helps to compare the experiments with the numerical simulations on the case of the reference package, but also serves at the same time for preliminary simulations with the package design to be approved.&#13;
The thermal test consists of a 30-minute fully engulfing 800°C pool fire or an equally severe fire, e.g. a propane gas fire. The fire reference test is performed prior to the regulatory fire test with the package design to be approved. The fire reference package in the described case is a closed sheet 316L steel cylinder with a wall thickness of 10 mm, a length of 182 mm, and a diameter of 102 mm. The package was instrumented with thermocouples and filled with heat resistant insulating material.</abstract>
    <enrichment key="eventName">ASME PVP 2024</enrichment>
    <enrichment key="eventPlace">Bellevue, WA, USA</enrichment>
    <enrichment key="eventStart">28.07.2024</enrichment>
    <enrichment key="eventEnd">02.08.2024</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>Tobias Gleim</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Experimental</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Numerical</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Heat Flux</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Reference Test</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="">Präsentation</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>60851</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
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    <pageNumber/>
    <edition/>
    <issue/>
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    <type>lecture</type>
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    <title language="eng">Safety Aspects for Transport Containers with regard to Ageing and Battery Accident Scenarios</title>
    <abstract language="eng">In Germany, the search for a final repository for highly radioactive material is currently based on the assumption that interim storage will take considerably longer than was anticipated. At that time, authorization procedures assumed that interim storage would be brief, until the containment systems could be placed in a final repository. For today's new approvals and extensions, the IAEA (International Atomic Energy Agency) regulations for the transport of radioactive material require an assessment of all components and component groups with regard to ageing. To assess the safety requirements not only over the previously envisaged periods but also over long durations, numerous individual parts and component groups must be examined and evaluated for ageing and interactions between materials. The metal seals of a double lid sealing system, for example, are critical components that must be studied with respect to ageing. In a containment system for radioactive material, metal seals must not only undergo classic ageing, but also withstand mechanical influences and radiation over extended periods. In addition to assessments by the applicants, BAM must also independently analyze and evaluate these ageing mechanisms as part of its sovereign mandate. &#13;
Another challenge in both new approvals and extensions is the assessment of transport regarding changes in the state of the art, which are not synchronously assessed in the regulations. In transporting radioactive material, a classic oil fire has been assumed in a hypothetical accident scenario from the development of the regulations to the present day. Due to the rapid development of propulsion technologies in recent years, such as battery or hydrogen-powered vehicles, investigations must be conducted in the revision process of the IAEA regulations to assess the impact of these new technologies on safety assessments. It is important for the member states of the IAEA to know whether new technologies are covered by the regulations or if changes are necessary to continue issuing international licenses. By focusing on batteries from the material to the cell to the module and the application, BAM contributes to the evaluation of batteries at every step of the chain. In addition to the safety of current and new battery technologies, BAM also endeavors to address issues of compositional trace and sustainable energy materials.</abstract>
    <enrichment key="eventName">MSD: Leaf</enrichment>
    <enrichment key="eventPlace">Livermore, CA , USA</enrichment>
    <enrichment key="eventStart">07.08.2024</enrichment>
    <enrichment key="eventEnd">07.08.2024</enrichment>
    <enrichment key="InvitedTalks">1</enrichment>
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    <author>Tobias Gleim</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Safety Aspects</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ageing</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Accident Scenarios</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Battery</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>60855</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst/>
    <pageLast/>
    <pageNumber/>
    <edition/>
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    <type>lecture</type>
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    <title language="eng">Experimental and Numerical Analyses for the Evaluation of Hydrogen as an Energy Source for Thermal Testing of Transport Packages of Radioactive Material</title>
    <abstract language="eng">In this paper we present a new hydrogen-based test rig for an ongoing feasibility study of using hydrogen as an energy source for the thermal testing of transport packages containing radioactive materials. The test rig will be capable of combusting hydrogen for a wide range of different burner geometries, mass flows and if necessary hydrogen blends. As this type of fire test according to the IAEA boundary conditions does not yet exist, a large number of preliminary investigations, safety assessments and calculations must be carried out in order to develop a viable concept for hydrogen fires. In the first step of the feasibility study, the temperature, structure, and radiation of various hydrogen flames are surveyed. In future works, the results will make it possible to design burner frames that are suitable for fire reference tests in order to make comparisons with pool and propane fires used in assessment procedures today. In parallel comparative numerical simulations are conducted to model the thermal behaviour of hydrogen flames using the software package Ansys®. On the one hand, the numerical simulations support the experiments by providing an overview of numerous parameters and the measuring range; on the other hand, they will help with the design of the burner frame in future work. This paper gives an overview in the design and capabilities of the test rig. Furthermore, the results of the parameter studies show that burner geometry and mass flow provide a significant design margin for the thermal shape of the hydrogen flames. In addition, the results of the initial numerical calculations will be used to determine the necessary sensors, the positions, and their operating range. Only the optimal interaction allows a controlled system that permits user-defined hydrogen fires.</abstract>
    <enrichment key="eventName">PVP2024, Pressure Vessels &amp; Piping Conference</enrichment>
    <enrichment key="eventPlace">Bellevue, Washington, USA</enrichment>
    <enrichment key="eventStart">29.07.2024</enrichment>
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    <author>Maximilian Naster</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Hydrogen</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Fire</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>IAEA Regulations</value>
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    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>CFD</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Burner</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>56134</id>
    <completedYear/>
    <publishedYear>2022</publishedYear>
    <thesisYearAccepted/>
    <language>deu</language>
    <pageFirst>15</pageFirst>
    <pageLast>20</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName>Fachverband für Strahlenschutz e.V.</publisherName>
    <publisherPlace>Berlin</publisherPlace>
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    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
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    <title language="deu">Radioaktive Stoffe in besonderer Form-wichtige Aspekte im Zulassungsverfahren</title>
    <abstract language="deu">Für die Beförderung von umschlossenen radioaktiven Stoffen (Strahler) können Transporterleichterungen gelten, wenn das Strahlerdesign nachweislich unfallsicher ausgelegt ist und eine Zulassung als radioaktiver Stoff in besonderer Form vorliegt. Die in Deutschland zuständige Behörde für die Prüfung und Zulassung radioaktiver Stoffe in besonderer Form ist die Bundesanstalt für Materialforschung und –prüfung (BAM).&#13;
Eine oft langfristige Nutzung der Strahler kann eine alterungsbedingte Schwächung der Umschließung des radioaktiven Stoffes zur Folge haben. Jeder Strahler einer zugelassenen Bauart muss jedoch im Beförderungsfall zu jeder Zeit, auch nach längerer Nutzung, den vorgeschriebenen Prüfanforderungen genügen. Die Alterungsbewertung auf der Grundlage einer vom Antragsteller zu spezifizierenden Nutzungsdauer des radioaktiven Stoffes in besonderer Form ist seit langer Zeit Bestandteil des Zulassungsverfahrens in Deutschland. Ein von Deutschland eingebrachter Vorschlag im Rahmen der anstehenden Revision der IAEA Empfehlungen für die sichere Beförderung von radioaktiven Stoffen (SSR-6) soll die Lesbarkeit des Regelwerkes bezüglich der Alterungsbewertung verbessern, Klarheit über die damit verbundenen Anforderungen erzeugen und damit zu einer Harmonisierung der Zulassungsverfahren auf internationaler Ebene beitragen. In diesem Beitrag sollen wichtige Einflussfaktoren auf das Alterungsverhalten von radioaktiven Stoffen in besonderer Form aufgezeigt und die Notwendigkeit für die Spezifikation einer Nutzungsdauer als Grundlage für die Alterungsbewertung begründet werden. Der deutsche Vorschlag für die Revision des IAEA Regelwerks zum sicheren Transport radioaktiver Stoffe (SSR-6) wird vorgestellt und erläutert.</abstract>
    <parentTitle language="deu">Strahlenschutz von den Grundlagen zur Praxis</parentTitle>
    <identifier type="issn">1013-4506</identifier>
    <enrichment key="eventName">Jahrestagung des FS 2022</enrichment>
    <enrichment key="eventPlace">Konstanz, Germany</enrichment>
    <enrichment key="eventStart">26.09.2022</enrichment>
    <enrichment key="eventEnd">30.09.2022</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <author>Annette Rolle</author>
    <author>Frank Wille</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Strahlenschutz</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Radioaktive Stoffe</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Beförderung</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Zulassung</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
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  </doc>
  <doc>
    <id>56137</id>
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    <publishedYear>2022</publishedYear>
    <thesisYearAccepted/>
    <language>deu</language>
    <pageFirst/>
    <pageLast/>
    <pageNumber/>
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    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="deu">Radioaktive Stoffe in besonderer Form - wichtige Aspekte im Zulassungsverfahren</title>
    <abstract language="deu">Für die Beförderung von umschlossenen radioaktiven Stoffen (Strahler) können Transporterleichterungen gelten, wenn das Strahlerdesign nachweislich unfallsicher ausgelegt ist und eine Zulassung als radioaktiver Stoff in besonderer Form vorliegt. Die in Deutschland zuständige Behörde für die Prüfung und Zulassung radioaktiver Stoffe in besonderer Form ist die Bundesanstalt für Materialforschung und –prüfung (BAM).&#13;
Eine oft langfristige Nutzung der Strahler kann eine alterungsbedingte Schwächung der Umschließung des radioaktiven Stoffes zur Folge haben. Jeder Strahler einer zugelassenen Bauart muss jedoch im Beförderungsfall zu jeder Zeit, auch nach längerer Nutzung, den vorgeschriebenen Prüfanforderungen genügen. Die Alterungsbewertung auf der Grundlage einer vom Antragsteller zu spezifizierenden Nutzungsdauer des radioaktiven Stoffes in besonderer Form ist seit langer Zeit Bestandteil des Zulassungsverfahrens in Deutschland. Ein von Deutschland eingebrachter Vorschlag im Rahmen der anstehenden Revision der IAEA Empfehlungen für die sichere Beförderung von radioaktiven Stoffen (SSR-6) soll die Lesbarkeit des Regelwerkes bezüglich der Alterungsbewertung verbessern, Klarheit über die damit verbundenen Anforderungen erzeugen und damit zu einer Harmonisierung der Zulassungsverfahren auf internationaler Ebene beitragen. In diesem Beitrag sollen wichtige Einflussfaktoren auf das Alterungsverhalten von radioaktiven Stoffen in besonderer Form aufgezeigt und die Notwendigkeit für die Spezifikation einer Nutzungsdauer als Grundlage für die Alterungsbewertung begründet werden. Der deutsche Vorschlag für die Revision des IAEA Regelwerks zum sicheren Transport radioaktiver Stoffe (SSR-6) wird vorgestellt und erläutert.</abstract>
    <enrichment key="eventName">Jahrestagung des FS 2022</enrichment>
    <enrichment key="eventPlace">Konstanz, Germany</enrichment>
    <enrichment key="eventStart">26.09.2022</enrichment>
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    <author>Annette Rolle</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Strahlenschutz</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Radioaktive Stoffe</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Beförderung</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Alterung</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Zulassung</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
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    <collection role="institutes" number="">3.0 Abteilungsleitung und andere</collection>
  </doc>
  <doc>
    <id>56141</id>
    <completedYear/>
    <publishedYear>2022</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst/>
    <pageLast/>
    <pageNumber/>
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    <title language="eng">Safety assessment of aged metal seals</title>
    <abstract language="deu">With our experiments we want to simulate a load situation which can result from a cask vertical drop test.  The lid can lift a little for a very short moment and that can perhaps result in a little movement of the seal, a little rotation or a little lateral movement, so that the contact section can change. The aim of the project is to learn more about the sealing efficiency after the seal is compressed again. What we know from former component tests with non-aged seals is, that the leakage rate measured after one or more decompression /compression cycles with seal movement becomes significant higher. And this result is considered for specification of design leakage rates for release calculation in safety cases. &#13;
In the recent years we all have learned more about the significant influence of temperature and time on seal behavior and so the plan was to do similar experiments with aged seals, just to get an idea about seal behavior and achievable leakage rates. &#13;
What we did in detail is:&#13;
We compressed Helicoflex-seals of both design types, with Aluminum and Silver outer jackets, in testflanges and aged them in an oven at a temperature of 125°C for 3 months to produce seal properties comparable with properties after several years use at more typical operating temperatures between 90 and 100°C.&#13;
After this ageing procedure we opened the flanges completely, moved the seal a little to vary the contact area and compressed the flanges again. What we measured during all compression and decompression cycles was the load, the deformation and -as long as possible- the leakage rates.</abstract>
    <enrichment key="eventName">Workshop IRSN/BAM</enrichment>
    <enrichment key="eventPlace">Cadarache, France</enrichment>
    <enrichment key="eventStart">12.10.2022</enrichment>
    <enrichment key="eventEnd">14.10.2022</enrichment>
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    <author>Annette Rolle</author>
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      <language>eng</language>
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      <value>Transport</value>
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      <language>eng</language>
      <type>uncontrolled</type>
      <value>Radioactive</value>
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    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Seal behavior</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ageing</value>
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  </doc>
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    <id>61353</id>
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    <publishedYear>2024</publishedYear>
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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</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 e</abstract>
    <enrichment key="eventName">ASME PVP 2024</enrichment>
    <enrichment key="eventPlace">Bellevue, WA, USA</enrichment>
    <enrichment key="eventStart">28.07.2024</enrichment>
    <enrichment key="eventEnd">02.08.2024</enrichment>
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    <author>Frank Wille</author>
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      <language>eng</language>
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      <value>Tightness verification</value>
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    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Weld seams</value>
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      <language>eng</language>
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      <value>Drop tests</value>
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    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Encapsulations for damaged spent nuclear fuel</value>
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    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Helium leak tests</value>
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