<?xml version="1.0" encoding="utf-8"?>
<export-example>
  <doc>
    <id>30135</id>
    <completedYear/>
    <publishedYear>2013</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>11</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName>Omnipress</publisherName>
    <publisherPlace/>
    <creatingCorporation>Institute of Nuclear Materials Management (INMM)</creatingCorporation>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Nuclear power plant waste management strategy in Germany</title>
    <abstract language="eng">The shut-down of eight nuclear power plants and the phase out of the nuclear energy genera� tion in Germany, due to a decision of the German government after the Fukushima accident, has rapidly forced the need for transportation and storage of spent nuclear fuel (SNF) from shut-down reactors in this country. In accordance with the present strategy in Germany, which &#13;
is characterized by searching for appropriate final repository areas, only the dry interim stor� age facilities are available for storage nuclear waste of the decommissioned plants as well as the operating plants. Therefore, the transportation and the storage casks must be available for both types of light water reactors. But for the time being this is the bottleneck. Unfortunately the prognoses are neglecting the fact, that such a high number of casks are needed in a very short space of time. This requires the acceleration of formal activeness in relation to the li� censes but also an increase of the production. In the meanwhile to get this process started, the only possibility is the storage in the pools at the plant site. Anyway, for the time being, this is one of the problems for dismantling the eight plants which have been shut down due to the decision of the Government, which is now fixed in the last amended German Atomic Energy Act dated 24th of February 2012. Starting to dismantle in the radioactive regions of the plants requires empty pools. For emptying the pools of damaged SNF special procedures regarding the transportation and storage are necessary. The transportation of damaged SNF requires special casks or at least special procedures and therefore also a special licenses. &#13;
This paper describes the challenge for the management in transportation and storage of SNF  in Germany after the shut-down of the nuclear power plants by reason of the Fukushima acci� dent.</abstract>
    <parentTitle language="eng">PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials (Proceedings)</parentTitle>
    <identifier type="old">33086</identifier>
    <enrichment key="eventName">PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials</enrichment>
    <enrichment key="eventPlace">San Francisco, CA, USA</enrichment>
    <enrichment key="eventStart">18.08.2013</enrichment>
    <enrichment key="eventEnd">23.08.2013</enrichment>
    <author>Anton Erhard</author>
    <author>Holger Völzke</author>
    <author>Bernhard Droste</author>
    <author>Dietmar Wolff</author>
    <author>Frank Wille</author>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
  </doc>
  <doc>
    <id>30136</id>
    <completedYear/>
    <publishedYear>2013</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>13</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName>Omnipress</publisherName>
    <publisherPlace/>
    <creatingCorporation>Institute of Nuclear Materials Management (INMM)</creatingCorporation>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Long-term performance of metal seals for transport and storage casks</title>
    <abstract language="eng">Dual purpose casks for the transportation and storage of spent nuclear fuel and other radioactive materials require very high leak-tightness of lid closure Systems under accident conditions as well as in the long-term to prevent activity release. For that purpose metal seals of specific types with an inner helical spring and outer metal liners are widely used and have shown their excellent performance if certain quality assurance requirements for fabrication and assembling are satisfied. Well defined surface roughness, clean and dry inert conditions are therefore essential. No seal failure in a loaded cask happened under these conditions until today.&#13;
Nevertheless, the considered and licensed Operation period is limited and all safety assessments have been performed and approved for this period of time which is 40 years in Germany so far. But in the meantime longer storage periods might be necessary for the future and therefore additional material data will be required. BAM is involved in the qualification and evaluation procedures of those seals from the early beginning. Because long-term tests are always time consuming BAM has early decided to perform additional tests with specific test seal configurations to gain a better understanding of the long-term behavior with regard to seal pressure force, leakage rate and useable resilience which is safety relevant mainly in case of accidental mechanical loads inside a storage facility or during a subsequent transport. Main test parameters are the material of the outer seal jacket (silver or aluminum) and the temperature. This paper presents the BAM test program including an innovative test mock-up and most recent test results. Based on these data extrapolation models to extended time periods are discussed, and also future plans to continue tests and to investigate seal behavior for additional test parameters are explained.</abstract>
    <parentTitle language="eng">PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials (Proceedings)</parentTitle>
    <identifier type="old">33087</identifier>
    <enrichment key="eventName">PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials</enrichment>
    <enrichment key="eventPlace">San Francisco, CA, USA</enrichment>
    <enrichment key="eventStart">18.08.2013</enrichment>
    <enrichment key="eventEnd">23.08.2013</enrichment>
    <author>Holger Völzke</author>
    <author>Dietmar Wolff</author>
    <author>Ulrich Probst</author>
    <author>Sven Nagelschmidt</author>
    <author>Sebastian Schulz</author>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
  </doc>
  <doc>
    <id>30159</id>
    <completedYear/>
    <publishedYear>2013</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>9</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName>Omnipress</publisherName>
    <publisherPlace/>
    <creatingCorporation>Institute of Nuclear Materials Management (INMM)</creatingCorporation>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">German approach and experience feedback of transport ability of SNF packages after interim storage</title>
    <abstract language="eng">In Germany the concept of dry interim storage of spent fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years.&#13;
The casks being used have an approved package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the recently revised “Guidelines for dry interim storage of irradiated fuel assemblies and heat-generating radioactive waste in casks” by the German Waste management Commission (ESK).&#13;
For transport on public routes between or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfil the specifications of the transport approval or other sufficient properties which satisfy the proofs for the compliance of the safety objectives at that time.&#13;
In recent years the validation period of transport approval certificates for manufactured, loaded and stored packages were discussed among authorities and applicants. A case dependant system of 3, 5 and 10 years was established. There are consequences for the safety cases in the Package Design Safety Report including evaluation of long term behavior of components and specific operating procedures of the package.&#13;
Present research and knowledge concerning the long term behavior of transport and storage cask components have to be consulted as well as experiences from interim cask storage operations. Challenges in the safety assessment are e.g. the behavior of aged metal and elastomeric gaskets under IAEA test conditions to ensure that the results of drop tests can be transferred to the compliance of the safety objectives at the time of transport after the interim storage period.</abstract>
    <parentTitle language="eng">PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials (Proceedings)</parentTitle>
    <identifier type="old">33117</identifier>
    <enrichment key="eventName">PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials</enrichment>
    <enrichment key="eventPlace">San Francisco, CA, USA</enrichment>
    <enrichment key="eventStart">18.08.2013</enrichment>
    <enrichment key="eventEnd">23.08.2013</enrichment>
    <author>Frank Wille</author>
    <author>Dietmar Wolff</author>
    <author>Bernhard Droste</author>
    <author>Holger Völzke</author>
    <author>M. Baden</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Storage</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Transport</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Lagerung</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Verpackung</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Radioactive material</value>
    </subject>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
  </doc>
  <doc>
    <id>33992</id>
    <completedYear/>
    <publishedYear>2015</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>8</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation>Nuclear institute (ni)</creatingCorporation>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Further research on cask impact limiter needed?</title>
    <abstract language="eng">Transport and storage casks for medium and high level radioactive waste are subjected to extreme heavy loads during the accidental drop scenarios prescribed by IAEA regulations and national storage acceptance criteria. Subsequently, considerable efforts have been made to optimize impact limiting structures aiming to reduce cask stresses and deformations. Though, their benefits can only fully be exploited, if the energy absorbing behavior is reliable simulated by numerical models which play an increasingly important role in safety assessments.&#13;
BAM has had conducted the 5-years research project ENREA in order to overcome the problems caused by the use of simplified, not sufficiently validated or even defective computational concepts for damping materials. While major results of the just recently completed program will be presented, the focus is laid on the question whether these Outputs provide a solid foundation for approximating impact limiting structures in all relevant configurations.&#13;
So far, material models have been implemented for spruce, Polyurethane foams and damping concrete based mainly on small and medium scale compression test results. The experimental program performed at BAM comprised likewise guided drop tests and, especially for concrete, penetration tests. Although the resulting numerical simulations showed good agreement with measured values, it turned out to be necessary to consider further effects with regard to e.g. multiaxial stress States, shearfailures of shock absorbing components and their interaction with enclosed or enclosing structures. Thus, the strengths and application constraints of the actual models will be shown as well as details about further research needs, if any.</abstract>
    <parentTitle language="eng">RAMTRANSPORT 2015 - International conference on the radioactive materials transport and storage (Proceedings)</parentTitle>
    <identifier type="old">37120</identifier>
    <enrichment key="eventName">RAMTRANSPORT 2015 - International conference on the radioactive materials transport and storage</enrichment>
    <enrichment key="eventPlace">Oxford, UK</enrichment>
    <enrichment key="eventStart">19.05.2015</enrichment>
    <enrichment key="eventEnd">21.05.2015</enrichment>
    <author>Eva Maria Kasparek</author>
    <author>Holger Völzke</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Numerical model</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Shock absorber</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Damping concrete</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Polyurethane foam</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Radioactive waste storage cask</value>
    </subject>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
  </doc>
  <doc>
    <id>35104</id>
    <completedYear/>
    <publishedYear>2015</publishedYear>
    <thesisYearAccepted/>
    <language>deu</language>
    <pageFirst>1</pageFirst>
    <pageLast>12</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>other</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="deu">Beitrag zu den Anforderungen an die Beschaffenheit von Endlagerbehältern hinsichtlich ihrer Rückholbarkeit und Bergbarkeit</title>
    <abstract language="deu">Die Bundesanstalt für Materialforschung und –prüfung (BAM) wurde von der AG 3 der Kommission Lagerung hoch radioaktiver Abfallstoffe gebeten, Anforderungen an Behälter für die Endlagerung hoch radioaktiver Abfallstoffe im Hinblick auf ih-re Rückholbarkeit während des Endlagerbetriebes und ihre Bergbarkeit nach Ver-schluss des Endlagers über einen Zeitraum von 500 (bis 1.000 ?) Jahren darzu-stellen. Die BAM kommt diesem Wunsch auf der Grundlage ihrer langjährigen Er-fahrungen und Fachkenntnisse auf den Gebieten der Bauartprüfung und der si-cherheitstechnischen Begutachtung von Transport- und Lagerbehältern für radio-aktive Stoffe nach.</abstract>
    <parentTitle language="deu">Kommission Lagerung hoch radioaktiver Abfallstoffe gemäß § 3 Standortauswahlgesetz, Arbeitsgruppe 3 Entscheidungskriterien sowie Kriterien für Fehlerkorrekturen, 14. Sitzung am 24.11.2015 (K-Drs./AG3-47)</parentTitle>
    <identifier type="old">38257</identifier>
    <identifier type="url">https://www.bundestag.de/blob/396206/f0a6276f4e2741c3022afb8ce64c0689/drs_047-data.pdf</identifier>
    <author>Holger Völzke</author>
    <author>Dietmar Wolff</author>
    <author>Eva Maria Kasparek</author>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
  </doc>
  <doc>
    <id>53114</id>
    <completedYear/>
    <publishedYear>2021</publishedYear>
    <thesisYearAccepted/>
    <language>deu</language>
    <pageFirst>1</pageFirst>
    <pageLast>10</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation>TU Bergakademie Freiberg</creatingCorporation>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="deu">Anforderungen an Behälter für die Endlagerung hochradioaktiver Abfälle in Steinsalz, Tonstein und Kristallingestein</title>
    <abstract language="deu">Bei der Auswahl eines Standortes für ein Endlager für hochradioaktive Abfälle (Wärme entwickelnde radioaktive Abfälle und ausgediente Brennelemente) in Deutschland werden auf Grundlage des Standortauswahlgesetzes von 2017 die drei potenziellen Wirtsgesteine Steinsalz, Tonstein und Kristallingestein in Betracht gezogen. Die Auswahl eines konkreten Endlagerstandortes steht dabei in engem Zusammenhang mit den für die jeweiligen Wirtsgesteine geeignet erscheinender Endlagersystemen. Zentraler Bestandteil dieser Endlagersysteme sind die Endlagerbehälter, deren Gestaltung und Auslegung maßgeblich die Randbedingungen für die Transport- und Einlagerungstechnik vorgibt und zudem die Nachweise zur Betriebs- und Langzeitsicherheit entscheidend prägt. Daher ist es für die Behälterentwicklung erforderlich, die an Endlagerbehälter zu stellenden Anforderungen systematisch herzuleiten. Auf dieser Grundlage können anforderungsgerechte Endlagerbehälter zielgerichtet für die jeweils zu untersuchenden Endlagerstandorte in unterschiedlichen Wirtsgesteinen entwickelt werden.&#13;
Dieser Beitrag fasst die wesentlichen Ergebnisse des von BAM und BGE TECHNOLOGY GmbH im Zeitraum 2017 bis 2020 bearbeiteten und vom Bundesministerium für Wirtschaft und Energie (BMWi) geförderten Forschungsvorhabens KoBrA zusammen, die in einem Abschlussbericht und vier zugehörigen Teilberichten umfassend dokumentiert sind.</abstract>
    <parentTitle language="deu">Proceedings Tage der Standortauswahl 2021</parentTitle>
    <enrichment key="eventName">Tage der Standortauswahl</enrichment>
    <enrichment key="eventPlace">Freiberg, Germany</enrichment>
    <enrichment key="eventStart">11.02.2021</enrichment>
    <enrichment key="eventEnd">12.02.2021</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Holger Völzke</author>
    <author>Dietmar Wolff</author>
    <author>Teresa Orellana Pérez</author>
    <author>Christian Herold</author>
    <author>W. Bollingerfehr</author>
    <author>A. Wunderlich</author>
    <author>S. Prignitz</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Endlager</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Behälter</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Sicherheitsanforderungen</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>hochradioaktive Abfälle</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>53119</id>
    <completedYear/>
    <publishedYear>2020</publishedYear>
    <thesisYearAccepted/>
    <language>deu</language>
    <pageFirst>1</pageFirst>
    <pageLast>254</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>report</type>
    <publisherName>BGE Technology</publisherName>
    <publisherPlace>Peine</publisherPlace>
    <creatingCorporation>Bundesanstalt für Materialforschung und -prüfung (BAM)</creatingCorporation>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="deu">Herleitung und Zusammenstellung der Anforderungen an Endlagerbehälter für ein HAW-Endlager in den drei  potenziellen Wirtsgesteinen Steinsalz, Tonstein und  Kristallingestein; Bericht zum Arbeitspaket 3 des FuE Verbundvorhabens KoBrA</title>
    <abstract language="deu">Der Bericht zum Arbeitspaket 3 baut auf den Ergebnissen der internationalen Recherche (AP1) sowie der Zusammenstellung der Randbedingungen und Beanspruchungsgrößen für die Endlagerbehälter (AP2) auf. Die Anforderungen an die Behälter – regulatorische, betriebliche sowie solche aus der Standortgeologie und den Einwirkungen und Prozessen, denen die Behälter ausgesetzt sind – werden hergeleitet und systematisch dargestellt. &#13;
Die für das Endlagersystem zu betrachtenden Nutzungsphasen werden aus den regulatorischen Anforderungen hergeleitet und charakterisiert. Die Einwirkungen auf die Behälter in den drei potenziellen Wirtsgesteinen sowie die sich für die Erfüllung der Sicherheitsfunktionen des Behälters als Teil des Endlagersystems ergebenden Behälterfunktionen werden ebenfalls wirtsgesteins- und zeitphasenabhängig betrachtet. Schließlich werden aus den Einwirkungen auf die Behälter und den Anforderungen an die Behälterfunktionen auch quantifizierbare Behälterfunktionen für alle drei potenziellen Wirtsgesteine abgeleitet.</abstract>
    <identifier type="url">https://www.bge-technology.de/fileadmin/user_upload/MEDIATHEK/f_e_berichte/KoBrA/2020-KoBrA_Bericht_AP3.pdf</identifier>
    <note>Auftraggeber: Bundesministerium für Wirtschaft und Energie (BMWi), Förderkennzeichen: 02E11527 und 02E11537.</note>
    <enrichment key="RelatedIdentifier">https://opus4.kobv.de/opus4-bam/frontdoor/index/index/docId/52289</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="RelatedIdentifier">https://opus4.kobv.de/opus4-bam/frontdoor/index/index/docId/53117</enrichment>
    <enrichment key="RelatedIdentifier">https://opus4.kobv.de/opus4-bam/frontdoor/index/index/docId/53118</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Christian Herold</author>
    <author>Holger Völzke</author>
    <author>Dietmar Wolff</author>
    <author>Teresa Orellana Pérez</author>
    <author>W. Bollingerfehr</author>
    <author>A. Wunderlich</author>
    <author>S. Prignitz</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Endlager</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Behälter</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Sicherheitsanforderungen</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Hochradioaktive Abfälle</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>52289</id>
    <completedYear/>
    <publishedYear>2020</publishedYear>
    <thesisYearAccepted/>
    <language>deu</language>
    <pageFirst>1</pageFirst>
    <pageLast>110</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>report</type>
    <publisherName>BGE Technology</publisherName>
    <publisherPlace>Peine</publisherPlace>
    <creatingCorporation>Bundesanstalt für Materialforschung und -prüfung (BAM)</creatingCorporation>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="deu">Abschlussbericht zum FuE-Verbundvorhaben KoBrA: Anforderungen und Konzepte für Behälter zur Endlagerung von Wärme entwickelnden radioaktiven Abfällen und ausgedienten Brennelementen in Steinsalz, Tonstein und Kristallingestein</title>
    <abstract language="deu">Im Forschungsvorhaben KoBrA werden Anforderungen an Endlagerbehälter für hochradioaktive Abfälle erstmals systematisch und unter Berücksichtigung des internationalen Standes von Wissenschaft und Technik hergeleitet. Parallel werden die behälterrelevanten Randbedingungen und Beanspruchungsgrößen für Endlagerbehälter in den drei potenziellen Wirtsgesteinen Steinsalz, Tonstein und Kristallingestein ermittelt. Darauf aufbauend werden die zur Erfüllung der Anforderungen unter den gegebenen Einwirkungen und Beanspruchungen notwendigen Behälterfunktionen bestimmt und erste Konzeptideen für zukünftige Endlagerbehälter diskutiert bzw. entwickelt.</abstract>
    <identifier type="doi">10.2314/KXP:1750909669</identifier>
    <note>Auftraggeber: Bundesministerium für Wirtschaft und Energie (BMWi), Förderkennzeichen: 02E11527 und 02E11537.</note>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="RelatedIdentifier">https://opus4.kobv.de/opus4-bam/frontdoor/index/index/docId/53117</enrichment>
    <enrichment key="RelatedIdentifier">https://opus4.kobv.de/opus4-bam/frontdoor/index/index/docId/53118</enrichment>
    <enrichment key="RelatedIdentifier">https://opus4.kobv.de/opus4-bam/frontdoor/index/index/docId/53119</enrichment>
    <author>Holger Völzke</author>
    <author>Christian Herold</author>
    <author>Dietmar Wolff</author>
    <author>Teresa Orellana Pérez</author>
    <author>W. Bollingerfehr</author>
    <author>S. Prignitz</author>
    <author>A. Wunderlich</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Endlager</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Behälter</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Anforderungen</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Hochradioaktive Abfälle</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Wirtsgestein</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>53372</id>
    <completedYear/>
    <publishedYear>2021</publishedYear>
    <thesisYearAccepted/>
    <language>deu</language>
    <pageFirst>1</pageFirst>
    <pageLast>10</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName/>
    <publisherPlace>Dresden</publisherPlace>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="deu">Aufbau eines Aufprallfundamentes mit reproduzierbaren Eigenschaften für Baumusterprüfungen von Abfallbehältern für das Endlager Konrad</title>
    <abstract language="deu">Abgeleitet aus den in Teil 3 der „Systemanalyse Konrad“ definierten Lastannahmen für den „untertägigen Anlagenbereich“ müssen Verpackungen, die der Abfallbehälterklasse I (ABK I) mit dem zusätzlichen Merkmal der störfallfesten (sf) Verpackung oder der ABK II zugeordnet werden, gemäß den „Endlagerungsbedingungen, Endlager Konrad“, nach einem „…Fall aus 5 m Höhe auf eine unnachgiebige Unterlage…“ bestimmten Anforderungen genügen.&#13;
Eine nähere Beschreibung der Prüfeinrichtung sowie die Definition detaillierter Anforderungen an das Aufprallfundament erfolgen in der „Produktkontrolle, Endlager Konrad“: „Die Prüfeinrichtung, bestehend aus einer Hebeeinrichtung und einem Fundament ist so zu gestalten, dass der Fall des Prüfmusters in jeder beliebigen Aufprallorientierung aus 5 m Höhe, gemessen von der Unterkante des Prüfmusters zum Fundament, möglich ist. Das Fundament ist aus Beton der Mindestgüte B 35 herzustellen und muss so beschaffen und verankert sein, dass keine Zerstörung in zahlreiche kleinere Bruchstücke erfolgt und keine horizontale Verschiebung des Fundamentes oder seiner Teile auftritt. Maße und Masse des Fundaments müssen so beschaffen sein, dass der Einfluss des Baugrundes unter dem Fundament auf den Aufprallwiderstand vernachlässigbar ist.“&#13;
Im Rahmen der Bauartprüfung für Behälter zur Endlagerung radioaktiver Abfälle mit zu vernachlässigender Wärmeentwicklung im Endlager Konrad wird die Bundesanstalt für Materialforschung und -prüfung (BAM) von der Bundesgesellschaft für Endlagerung (BGE) als Sachverständige hinzugezogen und ist in der Regel auch mit der Durchführung der notwendigen Fallprüfungen auf dem Testgelände Technische Sicherheit der BAM (BAM TTS) einschließlich der Bereitstellung der benötigten Prüfeinrichtungen sowie des regelkonformen Aufprallfundamentes beauftragt. Der Prozess, beginnend von der Gewährleistung einer von allen Beteiligten als gültig bewerteten Fallprüfung bis hin zur abschließenden Bewertung und Interpretation der Versuchsergebnisse erfordert dabei u.a. eine sorgfältige Planung und Fertigung der für das Endlager Konrad repräsentativen Fundamentplatte und deren Ankopplung an einen Untergrund im Sinne der „Regulations for the Safe Transport of Radioactive Material“ wie er zum Beispiel auf den beiden Fallversuchsanlagen des BAM Testgelände Technische Sicherheit (BAM TTS) zur Verfügung steht.&#13;
Um diese Anforderungen zu erfüllen muss zum einen der Erhalt der Integrität der Fundamentplatte beim Fallversuch durch eine geeignete Auswahl und kraftflussgerechte Anordnung der Stabstahl-Bewehrung sowie deren Verankerung garantiert und damit eine unzulässige Energieaufnahme durch Risse sowie eine unzulässige „… Zerstörung (der Fundamentplatte) in zahlreiche kleinere Bruchstücke …“, wie in [3] gefordert, verhindert werden. Außerdem ist zu gewährleisten, dass der Beton die Anforderungen an die vorgeschriebene Mindestgüte bzw. -festigkeit zum Zeitpunkt der Fallprüfung erfüllt, die Betonfestigkeit jedoch zur Vermeidung unnötig verschärfter Prüfrandbedingungen eine zu definierende Obergrenze nicht überschreitet.&#13;
Ausgehend von dem 2009 im Rahmen der KONTEC veröffentlichten Wissensstand präsentiert die BAM im vorliegenden Beitrag die wesentlichen Ergebnisse der zwischenzeitlich bei der Planung und Durchführung von zahlreichen Fallprüfungen gewonnenen Erfahrungen bzgl. Spezifikation und Fertigung der für die Fallversuche notwendigen Fundamentplatten sowie deren Anbindung an den Untergrund. Neben den immer wieder notwendigen Anstrengungen zur Gewährleistung einer im Rahmen der Definition korrekten Betondruckfestigkeit am Tag der Fallprüfung wird die Anpassung der Bewehrungsführung an verschiedene Fallpositionen diskutiert.&#13;
Zur Ergänzung des gültigen Regelwerks entsteht parallel zu diesem Dokument eine Fachnotiz, in der BGE und BAM die Vorgaben und Spezifikationen zur regelwerkskonformen Herstellung und Montage von Fundamentplatten für Fallprüfungen im Rahmen der Behälterbauartprüfungsverfahren für das Endlager Konrad zusammenfassen und veröffentlichen werden. Diese Fachnotiz soll detaillierte Vorgaben und Spezifikationen zur Planung, Beauftragung, Herstellung und abschließenden Prüfung eines solchen Fundamentes und dessen Anbindung an den Untergrund enthalten. Damit soll allen beteiligten Organisationen wie Antragstellern, Sachverständigen und BGE ein langfristig verlässlicher Leitfaden im Hinblick auf die Durchführung anforderungsgerechter Fallprüfungen unter definierten und reproduzierbaren Randbedingungen im Rahmen der Bauartprüfungsverfahren für das Endlager Konrad an die Hand gegeben werden.</abstract>
    <parentTitle language="deu">Tagungsband zur KONTEC 2021</parentTitle>
    <enrichment key="eventName">KONTEC 2021</enrichment>
    <enrichment key="eventPlace">Dresden, Germany</enrichment>
    <enrichment key="eventStart">25.08.2021</enrichment>
    <enrichment key="eventEnd">27.08.2021</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Mike Weber</author>
    <author>Holger Völzke</author>
    <author>Uwe Zencker</author>
    <author>Katharina Hauer</author>
    <author>A. Herz</author>
    <author>M. Kreienmeyer</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Aufprallfundament</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Endlager Konrad</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="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>55434</id>
    <completedYear/>
    <publishedYear>2022</publishedYear>
    <thesisYearAccepted/>
    <language>deu</language>
    <pageFirst>1</pageFirst>
    <pageLast>10</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="deu">Brittle Failure Analysis of High-Burnup PWR Fuel Cladding Alloys</title>
    <abstract language="deu">The general aim of this research is the development of methods for predicting mechanical behavior and identification of limiting conditions to prevent brittle failure of high-burnup (HBU) pressure water reactor (PWR) fuel cladding alloys. A finite element (FE) model of the ring compression test (RCT) was created to analyze the failure behavior of zirconium-based alloys with radial hydrides during the RCT. An elastic-plastic material model describes the zirconium alloy. The stress-strain curve needed for the elastic-plastic material model was derived by inverse finite element analyses. Cohesive zone modeling is used to reproduce sudden load drops during RCT loading. Based on the failure mechanism in non-irradiated ZIRLO® claddings, a micro-mechanical model was developed that distinguishes between brittle failure along hydrides and ductile failure of the zirconium matrix. Two different cohesive laws representing these types of failure are present in the same cohesive interface. The key differences between these constitutive laws are the cohesive strength, the stress at which damage initiates, and the cohesive energy, which is the damage energy dissipated by the cohesive zone. Statistically generated matrix-hydride distributions were mapped onto the cohesive elements and simulations with focus on the first load drop were performed. Computational results are in good agreement with the RCT results conducted on high-burnup M5® samples. It could be shown that crack initiation and propagation strongly depend on the specific configuration of hydrides and matrix material in the fracture area.</abstract>
    <parentTitle language="deu">SMiRT 26 Conference Proceedings</parentTitle>
    <enrichment key="eventName">26th International Conference on Structural Mechanics in Reactor Technology - SMiRT 26</enrichment>
    <enrichment key="eventPlace">Potsdam, Germany</enrichment>
    <enrichment key="eventStart">10.07.2022</enrichment>
    <enrichment key="eventEnd">15.07.2022</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Kai Simbruner</author>
    <author>M. C. Billone</author>
    <author>Uwe Zencker</author>
    <author>Y. Y. Liu</author>
    <author>Holger Völzke</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Cladding Embrittlement</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Cohesive Zone Modelling</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Spent Nuclear Fuel</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Ring Compression Test</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>59146</id>
    <completedYear/>
    <publishedYear>2023</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>8</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Fracture Mechanics Analysis of Spent Fuel Claddings during Long-Term Dry Interim Storage</title>
    <abstract language="eng">The prevention of brittle fracture of spent fuel claddings during long-term dry interim storage is based on experimental investigations, numerical analyses and assessment methods for predicting the mechanical behavior and determining limiting conditions. The ring compression test (RCT) is an established experimental method for characterizing cladding material. Test results for various high-burnup pressure water reactor zirconium-based fuel cladding alloys (e.g., ZIRLO®, M5®) are publicly available. To reduce the effort associated with irradiated samples in hot cells, it is helpful to perform studies on unirradiated surrogate cladding material. Based on such experimental data, load-displacement curves were numerically analyzed for selected cladding materials. In the presence of radial hydrides, a sample may suddenly fail by fracture even at small deformations. Noticeable load drops in the RCT occur associated to unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting neighboring hydrides. The failure process was simulated by cohesive zones controlled by the fracture energy and the cohesive strength. A modeling approach is presented in which the radial hydride morphology is taken into account. Based on the developed fracture mechanics approach with cohesive zone modeling, not only the deformation behavior but also the failure behavior of irradiated as well as unirradiated claddings with radial hydrides under RCT loading conditions can be adequately described.</abstract>
    <parentTitle language="eng">PATRAM 2022 Proceedings</parentTitle>
    <enrichment key="eventName">20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2022)</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>Uwe Zencker</author>
    <author>Mohan Reddy Gaddampally</author>
    <author>Holger Völzke</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Extended Interim Storage</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Nuclear Fuel Cladding</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Numerical Failure Analysis</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ring Compression Test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent Fuel Characterization</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>58563</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">Design of a Drop Test Target with Reproducible Properties for Konrad Prototype Testing</title>
    <abstract language="eng">Beginning in 1976 the former iron ore mine Konrad was geologically investigated concerning its suitability as a location for a German final repository for low and intermediate level waste. After a comprehensive licensing procedure it was approved and is now being prepared and equipped for the planned start of storage in 2027. At the end of the 1980s, the requirements for waste containers for storage in this repository were defined for the first time on the basis of an incident analysis of the on-site handling procedures, beginning with the unloading of the packages after the arrival at the site up to the final positioning in the underground storage galleries. A 5-meter drop onto the rock ground of the mine was identified as the covering case for high mechanical requirements (ABK II container class). In contrast to the 9-meter drop according to the IAEA Regulations for the Safe Transport of Radioactive Material, the 5-meter drop is performed onto a hard but not essentially unyielding target. The container is typically not protected by an impact limiter. The requirements for a potential test facility are described in the regulations for the Konrad repository. The mechanical strength of the target is defined as a concrete strength equivalent to the identified properties of the rock ground of the mine.&#13;
Since 1991 BAM has consistently used precast reinforced concrete slabs as target for drop tests in the framework of licensing procedures as well as in research projects. While the original design fulfilled the requirement for the integrity of the concrete slab in most cases, it failed when drop tests onto an edge of a container were performed. A redesigned concrete slab developed in a research project and suggested as a reference target in 2009 has been successfully used in Konrad licensing procedures since then.&#13;
The paper gives a brief overview of the historical development and it describes BAM’s efforts and approaches to continuously guarantee a concrete slab of defined quality and to provide a test setup for valid drop tests from the applicants and the authorities view.</abstract>
    <parentTitle language="eng">PATRAM 2022 Proceedings</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>Mike Weber</author>
    <author>Holger Völzke</author>
    <author>Uwe Zencker</author>
    <author>A. Herz</author>
    <author>M. Kreienmeyer</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Target</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Drop test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Final repository Konrad</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>56853</id>
    <completedYear/>
    <publishedYear>2022</publishedYear>
    <thesisYearAccepted/>
    <language>deu</language>
    <pageFirst>1</pageFirst>
    <pageLast>106</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>report</type>
    <publisherName>Bundesanstalt für Materialforschung und -prüfung (BAM)</publisherName>
    <publisherPlace>Berlin</publisherPlace>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="deu">Entwicklung eines bruchmechanischen Berechnungsansatzes zur Beschreibung des Festigkeitsverhaltens von Brennstabhüllrohren bei längerfristiger Zwischenlagerung (BRUZL)</title>
    <abstract language="deu">Es werden Methoden zur sicherheitstechnischen Bewertung des mechanischen Verhaltens von Brennstabhüllrohren entwickelt, um sprödes Versagen in Belastungsszenarien nach längerfristiger trockener Zwischenlagerung zu verhindern. Unbestrahlte Brennstabhüllrohrsegmente aus der Zirkoniumlegierung ZIRLO® wurden hydriert und thermomechanischen Behandlungen unterzogen, um Hüllrohrmaterial durch Ausscheidung radialer Hydride zu verspröden. Die Proben mit radialen Hydriden wurden im Ring Compression Test (RCT) bei Raumtemperatur untersucht. Bereits bei kleinen Probenverformungen wurden abrupte Kraftabfälle mit instabiler Rissausbreitung durch das Netzwerk radialer Hydride beobachtet. Die Risslängen erreichten in einigen Fällen bis zu 90 % der Wanddicke. Der mikromechanische Versagensmechanismus ist Quasi-Spaltbruch in den Hydriden und Porenbildung, -wachstum und -vereinigung in der Zirkoniummatrix an den duktilen Brücken zwischen benachbarten Hydriden. Es wurde ein Finite-Elemente-Modell des Ring Compression Tests erstellt, um das Versagensverhalten zirkoniumbasierter Legierungen mit radialen Hydriden im RCT zu analysieren. Die Fließkurve für das elastisch-plastische Materialmodell des Matrixmaterials wurde durch eine inverse Finite-Elemente-Analyse ermittelt. Kohäsivzonenmodelle wurden benutzt, um die plötzlichen Kraftabfälle zu reproduzieren. Es wurde ein mikromechanisches Modell mit statistisch generierten Matrix-Hydrid-Verteilungen entwickelt, das zwischen sprödem Versagen entlang der Hydride und duktilem Versagen der Zirkoniummatrix in der Kohäsivzone unterscheidet. Für das jeweilige Materialverhalten wird ein lokales zweiparametriges Versagenskriterium auf Grundlage der Kohäsionsfestigkeit und der Separationsenergie bei einer festgelegten Form des Kohäsivgesetzes vorgeschlagen. Geeignete Kohäsivparameter lassen sich mit einer inversen Finite-Elemente-Analyse des Versagensverhaltens von hydrierten Proben im Ring Compression Test bestimmen. In den Simulationen lag der Schwerpunkt auf dem ersten Kraftabfall. Die Berechnungsergebnisse stimmen gut mit den RCT-Ergebnissen überein. Es konnte gezeigt werden, dass die Rissinitiierung und -ausbreitung stark von der speziellen Anordnung der radialen Hydride und Matrixbrücken in der Bruchzone abhängt. Die numerische Modellbildung wurde an Versuchen mit bestrahlten Proben aus der Zirkoniumlegierung M5® validiert.</abstract>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Uwe Zencker</author>
    <author>Kai Simbruner</author>
    <author>Holger Völzke</author>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Zwischenlagerung</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Brennstabhüllrohr</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Versagensmechanismus</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Kohäsivzonenmodell</value>
    </subject>
    <subject>
      <language>deu</language>
      <type>uncontrolled</type>
      <value>Ring Compression Test</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>60323</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>10</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Brittle Failure Limits of Spent Fuel Claddings Subjected to Long-Term Dry Interim Storage Conditions</title>
    <abstract language="eng">The mechanical properties of spent fuel claddings can be adversely affected under the conditions of long-term dry interim storage, so that the failure limits may be reached in case of mechanical loads during handling or transport after storage. Pre-storage drying and the early stage of interim storage can expose the cladding to higher temperatures and higher tensile hoop stresses than those associated with in-reactor operation and pool storage. During slow cooling of a cladding tube under internal pressure, radial hydrides may precipitate in zirconium-based cladding alloys. This can lead to embrittlement of the material and sudden failure of the cladding integrity under mechanical stress. In order to prevent brittle failure, numerical methods are being developed to predict the mechanical behaviour and identify limiting conditions. Experimental investigations, numerical analyses and evaluation methods are discussed.&#13;
An established experimental method for characterising cladding materials is the Ring Compression Test (RCT). Some test results on irradiated cladding tubes after operation in pressure water reactors are publicly available. However, it is helpful to carry out studies on unirradiated surrogate claddings with similar material properties to reduce the effort associated with irradiated samples in hot cells and to perform material tests with a wider range of parameters. On the basis of such experimental data, load-displacement curves have been numerically analysed for a selection of cladding materials.&#13;
Radial hydrides can cause a sample to break suddenly due to fracture even at low deformation. Noticeable load drops in the RCT are caused by unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting adjacent hydrides. The cohesive zone approach was used to simulate the failure process taking into account the radial hydride morphology.&#13;
The developed method can adequately describe both the deformation and failure behaviour of irradiated as well as unirradiated claddings of zirconium-based alloys with radial hydrides under RCT conditions. Limiting conditions can be expressed in terms of fracture energy and cohesive strength.</abstract>
    <parentTitle language="eng">Proceedings of the IAEA International Conference on the Management of Spent Fuel from Nuclear Power Plants - Meeting the Moment</parentTitle>
    <enrichment key="eventName">IAEA International Conference on the Management of Spent Fuel from Nuclear Power Plants - Meeting the Moment</enrichment>
    <enrichment key="eventPlace">Vienna, Austria</enrichment>
    <enrichment key="eventStart">10.06.2024</enrichment>
    <enrichment key="eventEnd">14.06.2024</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Uwe Zencker</author>
    <author>Holger Völzke</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ageing Management</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cladding Embrittlement</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Extended Storage</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ring Compression Test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent Nuclear Fuel</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>60671</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>8</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Failure Analysis on Irradiated Claddings Subjected to Long-Term Dry Interim Storage</title>
    <abstract language="eng">Long-term dry interim storage may adversely affect the mechanical properties of spent fuel rods, possibly resulting in a reduced resilience during handling or transport after storage. The cladding is the first barrier for the spent fuel pellets. An established method for characterising the cladding material is the ring compression test (RCT), in which a small, cylindrical sample of the cladding tube is subjected to a compressive load. Radial hydrides may precipitate in zirconium-based alloys (Zircaloy) under pre-storage drying and during slow cooling, which result in embrittlement of the cladding material and eventually a possible sudden failure of cladding integrity under additional mechanical loads. The focus of the presented research is on the development of appropriate nu-merical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. A modelling approach based on cohesive zones is ex-plained which is able to reproduce the propagation of cracks initiated at radial hydrides in the zir-conium matrix. The developed methods are applied to defueled samples of cladding alloy ZIRLO®, which were subjected to a thermo-mechanical treatment to reorient existing circumferential hy-drides to radial hydrides. A selected sample showing sudden load drops during a quasi-static ring compression test is analysed by means of fracture mechanics for illustrative purposes. Based on the developed fracture mechanics approach, not only the deformation behaviour but also the fail-ure behaviour of irradiated as well as unirradiated Zircaloy claddings with radial hydrides under RCT loading conditions can be adequately described.</abstract>
    <parentTitle language="eng">SMiRT 27 Conference Proceedings</parentTitle>
    <enrichment key="eventName">27th International Conference on Structural Mechanics in Reactor Technology - SMiRT 27</enrichment>
    <enrichment key="eventPlace">Yokohama, Japan</enrichment>
    <enrichment key="eventStart">03.03.2024</enrichment>
    <enrichment key="eventEnd">08.03.2024</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="opus.doi.autoCreate">false</enrichment>
    <enrichment key="opus.urn.autoCreate">true</enrichment>
    <author>Uwe Zencker</author>
    <author>Mohan Reddy Gaddampally</author>
    <author>Holger Völzke</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cladding Embrittlement</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cohesive Zone Modelling</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Ring Compression Test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent Nuclear Fuel</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>44749</id>
    <completedYear/>
    <publishedYear>2017</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>8</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Relevance of cladding failure mechanisms due to zirconium hydride precipitation under extended dry storage conditions in Germany</title>
    <abstract language="eng">The German nuclear waste management strategy consists of dry interim storage of spent fuel inside dual purpose casks and its subsequent direct final disposal in a deep geological repository. Interim storage is limited to 40 years so far, but will have to be extended until repository site selection and operation is concluded, potentially up to 100 years. In this context, research on the long term performance of safety relevant cask components and fuel rod integrity during such extended periods of time is of outmost importance. The barrier function of the fuel cladding depends on its integrity under operational and accidental conditions along with storage and subsequent transportation and is mainly affected over time by altering of the cladding microstructure and by thermo-mechanical conditions during reactor operation and subsequent wet and dry storage. &#13;
This paper focuses on failure mechanisms regarding cladding embrittlement caused by physical phenomena concerning hydrogen dissolution and precipitation of hydrides in the zirconium matrix. Within this context, potential cladding integrity issues under extended interim storage conditions followed by transportation are emphasized.  &#13;
Delayed hydride cracking (DHC) is a well-known time-dependent temperature-activated phenomenon failure mechanism of the fuel cladding during reactor operation. The phenomenon consists of hydrogen diffusion and hydride precipitation at the tip of an initial crack and the propagation of the crack to an approximate distance equal to the length of the hydride. This failure mechanism is temperature-activated phenomenon and enhances along with thermal cycling of the fuel cladding, e. g. during reactor shutdowns. DHC shows limited relevance under quasi static storage conditions with continuously lowering temperatures. However, the literature points that DHC may occur when the temperature during storage decreases to a certain level after some decades and in combination with stress concentrations in the cladding material.&#13;
When spent fuel is transferred to dry storage by vacuum drying, the cladding temperature raises up to peak temperatures of ~370°C. Under these conditions, hydrides dissolute into the zirconium matrix up to the solubility limit defined by the peak temperature and this amount of hydrogen can precipitate in a radial re-oriented fashion due to the temperature decrease during storage. Cladding materials with radially precipitated hydrides show significant embrittlement. Radially oriented hydrides are perpendicularly oriented to hoop stresses representing the most critical configuration. Brittle failure of the cladding then occurs by crack propagation through radially oriented hydrides and finally through the zirconium matrix. The susceptibility to radial hydride precipitation depends on cladding material type and microstructure, hydrogen content, pre-drying hydride distribution, irradiation conditions, and temperature and stress histories during drying and storage operations.&#13;
This paper addresses major considerations concerning spent fuel cladding embrittlement due to hydride precipitation with regard to the boundary conditions of dry spent fuel storage in Germany. Analytical, numerical, and experimental approaches are to be discussed by BAM in order to identify the specific needs for future R&amp;D work in that area with the purpose to provide the necessary data base for proper safety demonstration and evaluation along with future extended storage licensing procedures.</abstract>
    <parentTitle language="eng">Proceedings of the 2017 Water Reactor Fuel Performance Meeting</parentTitle>
    <enrichment key="eventName">2017 Water Reactor Fuel Performance Meeting</enrichment>
    <enrichment key="eventPlace">Jeju, Korea</enrichment>
    <enrichment key="eventStart">10.09.2017</enrichment>
    <enrichment key="eventEnd">14.09.2017</enrichment>
    <author>Teresa Orellana Pérez</author>
    <author>Holger Völzke</author>
    <author>Dietmar Wolff</author>
    <author>Uwe Zencker</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cladding</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Hydride precipitation</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent fuel</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
  </doc>
  <doc>
    <id>43631</id>
    <completedYear/>
    <publishedYear>2017</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>Article UNSP V007T07A039, 1</pageFirst>
    <pageLast>9</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume>7</volume>
    <type>conferenceobject</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation>The Mechanical Society of Mechanical Engineers (ASME)</creatingCorporation>
    <contributingCorporation/>
    <belongsToBibliography>1</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Influence of impact angle and real target properties on drop test results of cubic containers</title>
    <abstract language="eng">Drop test scenarios with cubic containers without impact limiters at interim storage sites or in a final repository have been investigated by numerical simulations. An ideally flat drop is impossible to conduct as a free fall of a container even under laboratory conditions. Dynamic stresses and strains inside the container structure are sensitive to the impact angle. Even very small impact angles cause remarkable changes in the experimental or numerical results when a flat bottom or wall of a container hits a flat target. For drop tests with transport packages the International Atomic Energy Agency (IAEA) regulations define an essentially unyielding target. In contrast, potential accident scenarios for storage containers are derived from site-specific safety analyses or acceptance criteria in Germany. Each interim storage site or repository has a yielding or so-called real target with individual structural and material properties. The real target acts as a kind of impact limiter. A more conservative container design is required if the impact limiting effect of the target is not considered.</abstract>
    <parentTitle language="eng">Proc. ASME 2017 Pressure Vessels and Piping Conference</parentTitle>
    <identifier type="isbn">978-0-7918-5802-8</identifier>
    <identifier type="doi">10.1115/PVP2017-65731</identifier>
    <enrichment key="eventName">ASME 2017 Pressure Vessels and Piping Conference</enrichment>
    <enrichment key="eventPlace">Waikoloa, Hawaii, USA</enrichment>
    <enrichment key="eventStart">16.07.2017</enrichment>
    <enrichment key="eventEnd">20.07.2017</enrichment>
    <enrichment key="date_peer_review">08.02.2018</enrichment>
    <author>Uwe Zencker</author>
    <author>Linan Qiao</author>
    <author>Holger Völzke</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Finite element method</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Simulation</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Drop test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Impact angle</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Real target</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Yielding target</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
  </doc>
  <doc>
    <id>62558</id>
    <completedYear/>
    <publishedYear>2024</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>8</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume>5</volume>
    <type>conferenceobject</type>
    <publisherName>ASME</publisherName>
    <publisherPlace>New York City, USA</publisherPlace>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>1</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Consequences of an imperfectly mounted reinforcement cage in a generical cylindrical concrete container during mechanical specimen tests</title>
    <abstract language="eng">By highest court decision in 2007 the issued license for the Konrad repository (a former iron ore mine) for low and intermediate level radioactive waste was finally confirmed. Since then, the site is improved, prepared and equipped for the currently planned start of waste package emplacement by 2029. The Bundesanstalt fuer Materialforschung und -pruefung (BAM, Federal Institute for Materials Research and Testing) is regularly contracted by the Bundesgesellschaft für Endlagerung (BGE, Federal Company for Radioactive Waste Disposal) as the responsible authority for waste product control, waste container design testing, and waste package acceptance. In this context BAM is mainly responsible for design testing of various container types including the evaluation quality assurance measures for container manufacturing.&#13;
In the written Konrad disposal requirements all acceptance criteria for potential final disposal containers are defined. In addition to general requirements concerning container design, materials, dimensions and others, specific mechanical load scenarios the casks have to withstand are derived from an analysis of the on-site handling and emplacement procedures. Drop tests from different heights under consideration of the most damaging drop orientation and a stacking test are two examples for safety demonstrations that have to be performed during the container design testing procedure to qualify the respective container type.&#13;
The containers, whether cubic or cylindrical containers, can be made from different materials like ductile cast iron, reinforced concrete or as welded construction from steel plates and profiles. In case of concrete containers the reinforcement has the function to absorb tensile forces the concrete is not able to withstand by its own. Because of manufacturing imperfections of cylindrical concrete waste containers manufactured already decades ago, the reinforcement cages are not always exactly, axisymmetrically positioned when they were poured. This inspired BAM to perform various ABAQUS finite-element (FE) simulations by using the example of a simplified cylindrical container design with generic dimensions and pure elastic material properties. As outcome the stress conditions in the container during the drop test impact under different drop positions and during the stacking test were investigated and are explained and illustrated in this contribution. Different reinforcement cage positions were calculated and compared. On the basis of the calculated tensile and compressive stress distributions suggestions are derived on how a cylindrical concrete container with an imperfectly placed reinforcement cage has to be oriented for a drop or stacking test to ensure the most damaging test scenario.</abstract>
    <parentTitle language="eng">Conference Proceedings ASME 2024 Pressure Vessels &amp; Piping Conference</parentTitle>
    <identifier type="isbn">978-0-7918-8851-3</identifier>
    <identifier type="doi">10.1115/PVP2024-123797</identifier>
    <enrichment key="eventName">ASME 2024 Pressure Vessels &amp; Piping Conference</enrichment>
    <enrichment key="eventPlace">Bellevue, Washington, USA</enrichment>
    <enrichment key="eventStart">28.07.2024</enrichment>
    <enrichment key="eventEnd">02.08.2024</enrichment>
    <enrichment key="opus.source">publish</enrichment>
    <enrichment key="date_peer_review">26.05.2025</enrichment>
    <author>Mike Weber</author>
    <author>Holger Völzke</author>
    <author>Gregor Nieslony</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Container design testing procedure</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Reinforced concrete containers</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Drop test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Stacking test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Imperfectly mounted reinforcement cage</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Finite-element-analysis</value>
    </subject>
    <collection role="ddc" number="621">Angewandte Physik</collection>
    <collection role="institutes" number="">3 Gefahrgutumschließungen; Energiespeicher</collection>
    <collection role="institutes" number="">3.4 Sicherheit von Lagerbehältern</collection>
    <collection role="themenfelder" number="">Energie</collection>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
    <collection role="themenfelder" number="">Kerntechnische Entsorgung</collection>
  </doc>
  <doc>
    <id>33020</id>
    <completedYear/>
    <publishedYear>2015</publishedYear>
    <thesisYearAccepted/>
    <language>deu</language>
    <pageFirst>Paper 090, 786</pageFirst>
    <pageLast>794</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="deu">Präzisierung der Prüfanforderungen an Stahlblechcontainer für das Endlager Konrad</title>
    <abstract language="deu">Abfallgebinde, die im Endlager Konrad eingelagert werden sollen, müssen die Anforderungen und Kriterien erfüllen, die in den Endlagerungsbedingungen (SE-IB-29/08-Rev-1 [1]) und in der Produktkontrolle (SE-IB-30/08-Rev-1) festgelegt sind. Die Qualifizierung der Abfallbehälter erfolgt auf dieser Grundlage in Form von Bauartprüfungen, die von der Bundesanstalt für Materialforschung und -prüfung (BAM) im Auftrag des Bundesamtes für Strahlenschutz (BfS) durchgeführt werden. Darüber hinaus sind die im Hinblick auf Behälterbauartprüfungen relevanten Nebenbestimmungen aus dem Planfeststellungsbeschluss für das Endlager Konrad zu berücksichtigen. Zur besseren Anwendbarkeit werden diese Nebenbestimmungen in die Endlagerungsbedingungen und die Produktkontrolle eingearbeitet. Begleitend ist als Hilfestellung für die Abfallablieferer und Behälterhersteller ein Leitfaden geplant, der praktikable Nachweisstrategien aufzeigt und, soweit erforderlich, Hintergrundinformationen zu den Spezifikationen der Regelwerke liefert. Die BAM ist vom BfS beauftragt, hierfür Vorschläge zu entwickeln und Empfehlungen abzugeben.</abstract>
    <parentTitle language="deu">KONTEC 2015 - 12. Internationales Symposium "Konditionierung radioaktiver Betriebs- und Stilllegungsabfälle" (Proceedings)</parentTitle>
    <identifier type="old">36099</identifier>
    <enrichment key="eventName">KONTEC 2015 - 12. Internationales Symposium "Konditionierung radioaktiver Betriebs- und Stilllegungsabfälle"</enrichment>
    <enrichment key="eventPlace">Dresden, Germany</enrichment>
    <enrichment key="eventStart">25.03.2015</enrichment>
    <enrichment key="eventEnd">27.03.2015</enrichment>
    <author>Matthias Dittrich</author>
    <author>Eva Maria Kasparek</author>
    <author>Holger Völzke</author>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
  </doc>
  <doc>
    <id>32250</id>
    <completedYear/>
    <publishedYear>2014</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>T06 - Doc.183, 1-7</pageFirst>
    <pageLast/>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName/>
    <publisherPlace/>
    <creatingCorporation/>
    <contributingCorporation/>
    <belongsToBibliography>1</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">The German aging management approach for dry spent fuel storage in dual purpose casks</title>
    <abstract language="eng">Since the decision by the German government to face out nuclear electricity generation the total amount of spent nuclear fuel and high level wastes from reprocessing is limited and well determined. In addition the siting and licensing procedure to establish a final repository has been ruled by a new law in the mid of 2013 and further delays are very likely until a deep geological repository may start its operation. &#13;
&#13;
In the meantime dry interim storage in dual purpose casks being permanently certified for interim storage as well as transportation is the established technical solution. Several on-site as well former centralized facilities are operated successfully for many years but storage licenses are generally limited to 40 years and future lifetime extensions are predictable. Permanent aging management for storage facilities and casks is necessary to demonstrate compliance with safety requirements and furthermore to gain relevant data and information about the technical conditions of the facilities and their components for future lifetime extensions.&#13;
&#13;
For that reason procedures and measures are currently improved and the approach is explained in this paper. In addition, the current status and latest experiences concerning periodic safety inspections and aging management measures are discussed.</abstract>
    <parentTitle language="eng">PSAM 12 - Probabilistic safety assessment and management (Proceedings)</parentTitle>
    <identifier type="old">35316</identifier>
    <enrichment key="eventName">PSAM 12 - Probabilistic safety assessment and management</enrichment>
    <enrichment key="eventPlace">Honolulu, Hawaii, USA</enrichment>
    <enrichment key="eventStart">22.06.2014</enrichment>
    <enrichment key="eventEnd">27.06.2014</enrichment>
    <enrichment key="date_peer_review">15.12.2014</enrichment>
    <author>Holger Völzke</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Interim storage</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Dual purpose casks</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Spent nuclear fuel</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Aging management</value>
    </subject>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
  </doc>
  <doc>
    <id>30227</id>
    <completedYear/>
    <publishedYear>2013</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>Session H, Paper 202, 1</pageFirst>
    <pageLast>9</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName>Omnipress</publisherName>
    <publisherPlace/>
    <creatingCorporation>Institute of Nuclear Materials Management (INMM)</creatingCorporation>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Numerical analysis of cask accident scenarios in storage facilities</title>
    <abstract language="eng">Existing spent nuclear fuel (SF) and high active waste (HAW) management policies and practices worldwide are the result of past presumptions that sufficient reprocessing and/or disposal capacity would be available in the near term. Consequently, in the past many countries have developed specific solutions for different periods of time due to their individual national nuclear policies.&#13;
In Germany the concept of dry interim storage in dual purpose metal casks before disposal is being pursued for SF and HAW management and transport and storage licenses have been issued accordingly. The current operation licenses for existing storage facilities have been granted for a storage period of up to 40 years. This concept has demonstrated its suitability for over 20 years so far. Relevant safety requirements have been assessed for the short-term as well as for the long-term for site-specific operational and accidental storage conditions. But in the meantime significant delays in the national repository siting procedure occurred which will make extended storage periods necessary in the future.</abstract>
    <parentTitle language="eng">PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials (Proceedings)</parentTitle>
    <identifier type="old">33187</identifier>
    <enrichment key="eventName">PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials</enrichment>
    <enrichment key="eventPlace">San Francisco, CA, USA</enrichment>
    <enrichment key="eventStart">18.08.2013</enrichment>
    <enrichment key="eventEnd">23.08.2013</enrichment>
    <author>Mike Weber</author>
    <author>Uwe Zencker</author>
    <author>Dietmar Wolff</author>
    <author>Holger Völzke</author>
    <author>Uwe Herbrich</author>
    <author>Sven Nagelschmidt</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Numerical analysis</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Interim storage</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Cask accident scenario</value>
    </subject>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
  </doc>
  <doc>
    <id>30233</id>
    <completedYear/>
    <publishedYear>2013</publishedYear>
    <thesisYearAccepted/>
    <language>eng</language>
    <pageFirst>1</pageFirst>
    <pageLast>10</pageLast>
    <pageNumber/>
    <edition/>
    <issue/>
    <volume/>
    <type>conferenceobject</type>
    <publisherName>Omnipress</publisherName>
    <publisherPlace/>
    <creatingCorporation>Institute of Nuclear Materials Management (INMM)</creatingCorporation>
    <contributingCorporation/>
    <belongsToBibliography>0</belongsToBibliography>
    <completedDate>--</completedDate>
    <publishedDate>--</publishedDate>
    <thesisDateAccepted>--</thesisDateAccepted>
    <title language="eng">Model-sized and full-scale dynamic penetration tests on damping concrete</title>
    <abstract language="eng">Mechanical loading conditions of transport and storage casks for radioactive materials in accidental scenarios are highly affected by the behavior of both: the impact limiters and the footing materials. To minimize potential damages during the handling of casks, a so called damping concrete is frequently used for the footings in interim nuclear facilities. It obtains its shock absorbing properties through admixing of polymer cells. For a comprehensive mechanical evaluation of casks, advanced material models are also needed for damping concrete. In order to characterize the mechanical properties and to develop numerical material models, penetration tests were carried out at different test facilities of BAM. The tests contain static and dynamic penetration tests on cubic specimen with an edge length of 100 mm as well as mortared specimen with a size of 240 x 240 x 50 cm³. Indenters with different geometries and diameters were used for these model-sized penetration tests. Subsequently a full-scale cylindrical cast-iron indenter with a diameter of 110 cm was dropped of 5 m height on a realistic damping concrete footing.</abstract>
    <parentTitle language="eng">PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials (Proceedings)</parentTitle>
    <identifier type="old">33194</identifier>
    <enrichment key="eventName">PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials</enrichment>
    <enrichment key="eventPlace">San Francisco, CA, USA</enrichment>
    <enrichment key="eventStart">18.08.2013</enrichment>
    <enrichment key="eventEnd">23.08.2013</enrichment>
    <author>Robert Scheidemann</author>
    <author>Eva Maria Kasparek</author>
    <author>Karsten Müller</author>
    <author>Bernhard Droste</author>
    <author>Holger Völzke</author>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Drop test</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Dynamic</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Penetration</value>
    </subject>
    <subject>
      <language>eng</language>
      <type>uncontrolled</type>
      <value>Damping concrete</value>
    </subject>
    <collection role="fulltextaccess" number="">Datei im Netzwerk der BAM verfügbar ("Closed Access")</collection>
    <collection role="literaturgattung" number="">Graue Literatur</collection>
  </doc>
</export-example>
