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Eingeladener Vortrag
- nein (88)
Entsprechend der Zielsetzung des Vorhabens stand die weitere Etablierung des Werkstoffes Sphäroguß (GGG-40) im Bereich der Transport- und Lagerbehälter für radioaktive Stoffe im internationalen Rahmen im Vordergrund der durchgeführten Untersuchungen und Aktivitäten. Dabei gelang es, eine weitgehende fachliche Übereinstimmung hinsichtlich der Bewertungsmethoden und -kriterien zu erzielen und auf IAEA-Ebene zu etablieren. Zuvor vorhandene Vorbehalte insbesondere auf U.S.-amerikanischer Seite konnten wesentlich vermindert werden.
Im Rahmen der durchgeführten Untersuchungen konnte die außerordentliche Belastbarkeit eines Prototypbehälters aus Sphäroguß (CASTOR VHLW) unter extremen Beanspruchungssituationen nachgewiesen werden. So war selbst bei Spannungen im Bereich der Material-Streckgrenze und vorhandenem rißartigen Fehler an dieser Stelle keine Sprödbruchneigung des Werkstoffes erkennbar. Den entsprechenden 14 m-Fall auf zwei harte Widerlager mit daraus folgender hoher Biegebeanspruchung überstand der Behälter unbeschadet.
Mittels begleitender Werkstoffuntersuchungen wurden die Kenntnisse über die Zusammenhänge zwischen der chemischen Zusammensetzung der Gefügeausbildung und den mechanischen Eigenschaften von Sphäroguß erweitert. Dabei wurden sowohl unterschiedliche Gießverfahren (Sand- und Kokillenguß) in der Herstellung als auch die Auswirkungen von Umwandlungsprozessen im Gefüge infolge von Wärmebehandlungen diskutiert. Die statistische Auswertung von Einzelergebnissen im Rahmen der Qualitätssicherung bei der Begutachtung von Transport- und Lagerbehältern für radioaktive Stoffe wurde durch mechanische und bruchmechanische Untersuchungen an 24 Werkstoffsorten aus Gußeisen mit Kugelgraphit ergänzt.
Insgesamt konnten somit die Entwicklung und der Einsatz von Sphäroguß für Transport- und Lagerbehälter radioaktiver Stoffe durch umfangreiche experimentelle und analytische Untersuchungen sowie zahlreiche Fachdiskussionen auf internationaler Ebene positiv beeinflußt und wesentlich vorangetrieben werden.
Im Rahmen des Forschungsvorhabens wurde von der Bundesanstalt für Materialforschung und -prüfung ein Begutachtungskonzept auf bruchmechanischer Basis für sog. Typ B(U)-Behälter aus Sphäroguß zum Transport und zur Lagerung radioaktiver Stoffe entwickelt, siehe dazu Abschlußbericht und Fachband 4. Zur Verifizierung des erarbeiteten Konzeptes wurden sowohl Fallversuche an einem Prototypbehälter, siehe Fachband 2, als auch umfangreiche Werkstoffuntersuchungen durchgeführt, siehe Fachband 3. Um jedoch die experimentellen Untersuchungen sinnvoll aufeinander abzustimmen, bedurfte es der Erarbeitung der entsprechenden theoretischen Grundlagen, welche Gegenstand dieses Fachbandes sind. Dabei wird das für die Berechnung gewählte Behältermodell erläutert und werden die im Hinblick auf unterschiedliche Fallsituationen erzielten Berechnungsergebnisse diskutiert. Diese dienten u.a. dazu, bestimmte Versuchsergebnisse (z.B. maximale Spannungen am Behälter) im Voraus abzuschätzen, um die für die bruchmechanischen Untersuchungen erforderlichen Versuchsparameter zweckmäßig einzustellen. Weiterhin weist der vorliegende Fachband in knapper Form auf die wesentlichen Grundlagen der relevanten Bereiche der Bruchmechanik sowie der Sprödbruchanalyse hin. Insbesondere der Fachband 4 beinhaltet hierzu weitergehende Details, die im Rahmen internationaler Fachdiskussionen auf IAEA-Ebene unter Mitwirkung der BAM als Stand von Wissenschaft und Technik gelten können.
Ziel der im Rahmen dieses Forschungsvorhabens durchgeführten Untersuchungen war der Nachweis der sicheren Auslegung von Transport- und Lagerbehältern radioaktiver Stoffe aus Gußeisen mit Kugelgraphit auch unter bruchmechanischen Gesichtspunkten. Das bisher im nationalen Rahmen für die Bauartprüfung derartiger sogenannter Typ B-Verpackungen (d.h. unfallsicherer Verpackungen) von der BAM angewandte Begutachtungskonzept basiert im wesentlichen auf experimentellen Festigkeitsnachweisen an Prototypbehältern mit entsprechend daraus abgeleiteten Anforderungen an die qualitätsgerechte Herstellung der Serienbehälter. Die Anforderungen an die Werkstoffqualität und die zerstörungsfreien Prüfungen (insbesondere Nachweis von Rissen durch Ultraschallprüfungen) werden so gewählt, daß ein gesonderter bruchmechanischer Nachweis nicht erforderlich ist. Dieses Begutachtungskonzept ist jedoch unmittelbar mit der Bedingung verknüpft, daß die maximale in der Behälterstruktur auftretende Nennzugspannung etwa 50 % der Werkstoff-Streckgrenze nicht überschreitet. Da dies u.U. eine erhebliche Einschränkung für bestimmte Behälterbauarten bedeuten kann, wird das Begutachtungskonzept derzeit um einen bruchmechanischen Festigkeitsnachweis erweitert, der weitgehend werkstoffunabhängig auch für Gußeisen mit Kugelgraphit anwendbar ist. Die Bereitstellung der dafür erforderlichen grundlegenden Erkenntnisse wurde unter direkter Mitwirkung der BAM auch im Rahmen dieses Forschungsvorhabens auf internationaler Ebene durchgeführt. Die Ergebnisse sind in einem "Technical Dokument (TECDOC)" der Internationalen Atomenergie-Behörde (IAEA) niedergelegt, welches diesem Fachband als Anlage beigefügt ist.
Die sicherheitstechnische Auslegung von Behältern aus duktilem Gusseisen mit erhöhten Reststoffanteilen wird untersucht. Ausgehend von den zu gewährleistenden Anforderungen an Transport- und Lagerbehälter für radioaktive Abfälle werden Referenzbeanspruchungs- situationen definiert. Für einen ausgewählten Referenzfall (bestehend aus der Kombination der Behälterbauart Gusscontainer Typ VI, einem Absturz aus 5 m Höhe auf ein hartes Endlagergestein simulierendes Betonfundament als repräsentativer Störfall und vorgegebenen Werkstoffeigenschaften) wird eine numerische Spannungsanalyse durchgeführt. Die Berechnungen erfolgen mit der Methode der finiten Elemente an dreidimensionalen Modellen, deren numerische Stabilität und Genauigkeit überprüft wird. Zur Berechnung der Behälterbeanspruchungen in Abhängigkeit vom Reststoffanteil werden dynamische Fließkurven für definiert legierte Testabgüsse angegeben.
Die Koordinierungsstelle für Informationen zur Behälterabfertigung (KOBAF): Aufgaben und Erfahrungen
(2008)
BAM is the responsible authority in Germany for the assessment of the mechanical and thermal design safety of packages for the transport of radioactive materials. The assessment has to cover the proof of brittle fracture safety for package components made of potentially brittle materials. This paper gives a survey of the regulatory and technical requirements for such an assessment according to BAM's new 'Guidelines for the application of ductile cast iron for transport and storage casks for radioactive materials'. Based on these guidelines, higher stresses than before will be permissible, but it is necessary to put more effort into the safety assessment procedure. The fundamentals of such a proof using the methods of fracture mechanics are presented. The recommended procedure takes into account the guidelines of the IAEA's advisory material which are based on the prevention of crack initiation. Examples of BAM's research and safety assessment practices are given. Recommendations for further developments towards package designs with higher acceptable stress levels will conclude the paper.
Countdown für Konrad
(2010)
Casks for the transport and storage of heat generating radioactive waste in Germany are normally provided with screwed lid systems, which are in most cases equipped with double jacket metal seals with an inner spring wire to provide long term resistance to the seal compression force. Preservation of the high sealing quality of those seals under operational and accidental stress conditions is essentially important to the safety of those casks. Relative displacements of the lid system surfaces caused by specific impact scenarios cannot be excluded and have to be evaluated with respect to a possible increase in the leakage rate.
To get representative data for such metal sealed lid systems, BAM has developed a special conceptualised flange system placed in an appropriate testing machine for relevant mechanical loading of the metal seals under static and cyclic conditions. Furthermore, the flange system enables continuous measurement of the standard helium leakage rate during each test.
The primary aim of the investigation is to identify the correlation between variation of installation conditions (axial displacements) caused by external loads and the standard helium leakage rate. An essential parameter in this case is the useable resilience ru of a metal seal under relevant stress conditions. The useable resilience ru is the vertical difference in the cross-section between the seal's assembling status and the point where the leakage rate, by means of external load relieving, exceeds the quality criterion of 10-8 Pa m3 s-1. Load relieving can instantly occur due to modification of the seal groove dimension caused by accident impacts and deformation of the lid system. Furthermore, component specific basis data for the development of finite element calculation models should be collected. In the tests, seals are subjected to static and cyclic loads. All tests are performed at ambient temperature.
This paper presents the test configuration, different test series and results of the current experiments. Typical load-displacement-leakage rate correlations are presented and discussed.
Erfahrungen und Perspektiven bei der Bauartprüfung von Behältern zur Verpackung radioaktiver Abfälle
(2008)
Die sicherheitstechnische Prüfung von Tranport- und Lagerbehältern für abgebrannte Kernbrennstoffe
(1998)
Approval of existent waste packages and new package design in preparation for the KONRAD repository
(2008)
Mechanical and thermal safety analyses and demonstrations for cubic DCI multipurpose containers
(1997)
Mechanical and thermal safety analyses and demonstrations for cubic DCI - multi-purpose containers
(1997)
Strategien zur Bestimmung von Temperaturen von Transport- und Lagerbehältern für radioaktive Stoffe
(2010)
Proper predictions of the behaviour of shock absorber materials are of utmost importance in safety assessments for licensing casks for transport and storage of highly active waste. After having identified significant discrepancies between numerical results and the actual response of polyurethane foam limiters subjected to accidental scenarios, a new research project ENREA was established by BAM. A major objective is to enhance and to develop advanced material models intended to simulate limiters under impact loading. They should account for all major factors influencing the load–deformation relationship like temperature, strain rate and specific stress state. The corresponding test program, applicable plasticity models, the overall parameter identification strategy based on local and global optimization techniques as well as experimental and numerical results are presented here in particular for closed cell foams.
The Konrad repository for not heat generating radioactive wastes was licensed first in 2002. But due to legal actions this license has not been finally confirmed before 2007. Subsequently, the Federal Office for Radiation Protection (BfS) began scheduling backfitting of the former iron ore mine into a repository. The licensed repository volume is 303,000 m3 based on estimations of expected waste volumes to be disposed off although the mine itself would offer a much larger volume. Once the repository is ready for Operation, waste packages can be disposed, but this is not expected before the end of this decade. Nevertheless, there is already today a great interest in qualified and certified waste conditioning and packaging for disposal requiring Containers, tested, evaluated and certified by BAM and BfS.
Key points of the Konrad license concern comprehensive requirements for waste products and Containers. Details are described in BfS reports, which constitute the basic documents for the license. Following the final license confirmation a revision of these documents has been requested in order to include considerations of water poisoning substances. This work is currently done by BfS in Cooperation with the NLWKN of Lower Saxony and the Institute for Safety Technology (ISTec). Concerning the specification for Containers themselves no or only minor changes are expected.
In drop test scenarios related to assessing and licensing the storage procedure of spent fuel and high active waste, the casks under examination are generally not equipped with impact limiters. Hence, the extent of mechanical stresses in case of an assumed handling accident is largely affected by the ground properties of the reception hall floor in the specific storage facility.
Unlike conventional brittle foundation materials, damping concrete performs quite well in such applications as it features high stiffness as well as high energy absorption due to the filler pore volume. However, its damping ability is not sufficiently exploited in current finite element (FE) calculations due to a lack of advanced material models for simulating its impact response. An implementation of qualified concepts that account for plastic, strain rate dependent behavior requires additional information that has to be provided by systematic test series.
BAM recently started a research project to generate such data, subsequently to develop and to improve numerical methods for the analysis of impact limiters and damping foundation material and thus to optimize safety assessment tools for the design of transport and storage casks. A major part of this research concerns dynamic compression tests of variably shaped specimens conducted at a servo hydraulic 1MN impact testing machine as well as at a BAM facility for guided drop tests. This presentation focuses 100mm damping concrete cubes deformed vertically at constant rates under different constraint conditions. For example, a special fitting jig was constructed to subject the specimens to multi-axial loading. Thereby a deformation of 60% could be applied.
Simulation was conducted by FE code ABAQUS based on material models Concrete damaged plasticity and Crushable foam which both allow defining rate sensitive nonlinear stress-strain relations in compression beyond the classic metal plasticity approach.
Die mechanische Beanspruchung von Transport- und Lagerbehältern für radioaktive Stoffe gemäß den durch die IAEA vorgegebenen Prüfszenarien oder bei Unfällen, die während der Handhabung in einer kerntechnischen Anlage auftreten können, wird maßgeblich durch das Design der Stoßdämpfer und die Eigenschaften des jeweiligen Untergrundes beeinflusst. Da der Nachweis der Behälterintegrität und -dichtheit häufig an numerische Simulationen gekoppelt ist, sind für die sicherheitstechnische Auslegung der Konstruktion zuverlässige Informationen über das Energieabsorptionsvermögen der Werkstoffe und eine geeignete Umsetzung dieser Daten in FEM - Programme erforderlich. Die im Rahmen von Falluntersuchungen bei der BAM festgestellten, teilweise erheblichen Abweichungen zwischen experimentellen und berechneten Größen sind entsprechend darauf zurückzuführen, dass die zur Verfügung stehenden numerischen Verfahren das komplexe Verhalten der stoßdämpfenden Strukturen
nicht ausreichend präzise wiedergeben. Hiervon besonders betroffen sind Bauteile aus Holz und Polyurethan-Schaum (PU-Schaum), die zur Konstruktion von Behälterstoßdämpfern für die Transportkonfiguration eingesetzt werden, sowie Dämpferbeton, der in ausgewählten Bereichen im Fundament von Standort-Zwischenlagern eingebaut wird. Um die zugehörigen Materialmodelle weiter zu entwickeln und die dafür erforderliche Datenbasis zu schaffen, hat die BAM vor knapp zwei Jahren das vom BMBF geförderte Forschungsvorhaben ENREA (Entwicklung rechnerischer Analysemethoden für stoßdämpfende Strukturen beim Anprall oder Absturz von Abfallgebinden) gestartet. Wesentlicher Bestandteil ist ein systematisches Versuchsprogramm, bei dem die drei genannten Werkstoffe statisch und dynamisch hohen Druckbelastungen ausgesetzt werden. Relevante Versuchsparameter sind, neben der Dehnrate, die Temperatur und Lagerungsbedingungen sowie die Orientierung des Werkstoffes hinsichtlich der Belastungsrichtung. Dieser Beitrag präsentiert die wichtigsten Ergebnisse aus der ersten Versuchsphase, die durch weggesteuerte Versuche mit konstanten Prüfgeschwindigkeiten von 0,02 mm/s und 3000 mm/s gekennzeichnet ist. Darüber hinaus wird die Verwertung dieser Daten zur Anpassung von Materialmodellen auf der Grundlage von Nachrechnungen gezeigt, wobei die Möglichkeiten und Grenzen der verwendeten Konzepte Umrissen werden. Aufgrund des Umfangs, den die Versuche mit Holzwerkstoffen einnehmen, und des erforderlichen Aufwandes für die numerische Modellierung werden die entsprechenden Maßnahmen und Ergebnisse in einem eigenständigen Beitrag auf der KONTEC vorgestellt.
Im Zuge des Forschungsvorhabens ENREA (Entwicklung rechnerischer Analysemethoden für stoßdämpfende Strukturen beim Anprall oder Absturz von Abfallgebinden) führt die BAM systematische Untersuchungen an stoßdämpfenden Werkstoffen, unter anderem an Fichtenholz, durch. Zweck der mit diesen Untersuchungen erzeugten Datenbasis ist die Erstellung und Parametrierung eines Finite-Elemente- (FE-) Materialmodells, das die Modellierung holzgefüllter stoßdämpfender Bauteile von Transportbehältern für radioaktive Stoffe in der FE-Simulation ermöglichen soll.
Anhand von Ergebnissen aus der ersten Versuchsphase werden der Einfluss der Dehnrate und der Faserorientierung auf das Kraft-Verformungs-Verhalten von Fichtenholz diskutiert. Anschließend wird die rechnerische Modellierung von Fichtenholz diskutiert und die Anforderungen an ein Materialmodell werden erläutert.
Dry cask storage of spent fuel and high active waste front reprocessing is licensed and established in Germany for up to 40years so far. About 10years ago, the strategy has been shifted from centralized storage using facilities in Ahaus and Gorleben to cit-site storage facilities in Order to avoid immediate cask shipping from nuclear power plants. At present, several hundred casks are stored safeiy and without any major problem. BAM is highly involved in cask specific safety evaluations within the scope of licensing procedures as well as of quality assurance procedures during cask fabrication, loading and storage.
A major prerequisite for any cask storage license is a valid transport Type B license. Subsequently, cask transportability has to be ensured at any point during the entire storage period. So far, this requirement is met by continuous maintenance of the Type B license. Anyhow, IAEA Type B regulations do not consider permanently loaded storage casks that have the sole purpose of being finally shipped öfter decades of storage inducing ageing effects. No transport safety demonstration has to be performed on aged cask components.
Das Endlager Konrad wurde bereits im Jahr 2002 als Endlager für nicht Wärme entwickelnde radioaktive Abfälle genehmigt. Jedoch wurde diese Genehmigung anschließend beklagt und konnte erst 2007 in letzter Instanz bestätigt werden. Nachfolgend hat das Bundesamt für Strahlenschutz (BfS) mit der Planung und Realisierung zum Umbau des ehemaligen Eisenerzbergwerkes zum Endlager begonnen. Das genehmigte Endlagervolumen beträgt 303.000 m3 und basiert auf Abschätzungen für die zu erwartende Abfallmenge. Das Bergwerk selbst böte ein deutlich größeres Volumen. Mit der Einlagerung von Behältern kann allerdings erst nach Fertigstellung des Endlagers und nach derzeitigem Stand voraussichtlich nicht vor Ende dieses Jahrzehnts gerechnet werden. Gleichwohl besteht bereits heute ein hohes Interesse an der endlagergerechten Konditionierung und Verpackung der für Konrad vorgesehenen radioaktiven Abfälle, wofür nicht zuletzt von der BAM bauartgeprüfte und vom BfS zugelassene Behälter erforderlich sind.
Kernpunkte der Endlagergenehmigung sind umfassende Anforderungen an Abfallprodukte und Behälter. Einzelheiten sind in den vom BfS der Genehmigung zu Grunde liegenden Berichten und Endlageranforderungen definiert. Diese enthalten vor allem Anforderungen an die Abfallprodukte und an die zu deren Verpackung erforderlichen Behälter. Mit der Genehmigungserteilung wurde auch eine Überarbeitung dieser Endlageranforderungen und hinsichtlich der Berücksichtigung Wasser gefährdender Stoffe gefordert. Diese Arbeiten werden derzeit vom BfS in Zusammenarbeit mit dem Niedersächsischen Landesbetrieb für Wasserwirtschaft, Küsten- und Naturschutz (NLWKN) und dem Institut für Sicherheitstechnik (ISTec) vorgenommen. Hinsichtlich der Anforderungen an die Endlagerbehälter wird allerdings davon ausgegangen, dass die für die Behälterbauartprüfung wesentlichen abfallprodukt- und behälterspezifischen Anforderungen davon nicht oder nur unwesentlich berührt sein werden.
Während des Transportes sind Behälter für hoch radioaktive Abfälle durch stoßdämpfende Bauteile vor Einwirkungen, die aus Unfallszenarien gemäß den IAEA Regularien resultieren können, weitestgehend geschützt. Nach dem Entfernen dieser Komponenten im Rahmen der Einlagerung ins Zwischenlager müssen Vorkehrungen getroffen sein, um Beanspruchungen der Behälter aus möglichen Handhabungsunfällen im Empfangsbereich zu reduzieren. Die Maßnahmen müssen vor allem darauf abzielen, die stoßdämpfenden Eigenschaften der potentiellen Aufprallflächen zu optimieren. Da Standardmaterialien für Fundamente nur ein geringes Energieabsorptionsvermögen aufweisen, wurde durch die Firma HOCHTIEF eine neuartige Betonmixtur entwickelt, bei der ein spezieller Leichtzuschlag zum Einsatz kommen. Zur Charakterisierung dieses sogenannten Dämpferbetons liegen Standardkennwerte vor sowie Messwerte aus Eindringversuchen, die zur Qualitätssicherung für jede Produktcharge durchgeführt werden. Das stoßdämpfende Potential des neuen Werkstoffs kann jedoch nur dann vollkommen ausgeschöpft werden, wenn die Energieabsorption in numerischen Simulationen der Behälterabstürze über die gesamte Belastungsdauer zuverlässig wiedergegeben wird. Für die Implementierung und Anpassung eines entsprechenden Materialmodelis in einem Finite Elemente (FE) Programm
sind Daten aus systematischen Druckversuchen mit Variation der relevanten Einflussgrößen erforderlich. An der BAM werden daher im Rahmen eines werkstoffübergreifenden Forschungsvorhabens zur Untersuchung von stoßdämpfenden Strukturen auch die Eigenschaften von Dämpferbeton experimentell untersucht und zugehörige numerische Methoden entwickelt. Der aktuelle Beitrag behandelt die Ergebnisse der ersten Projektphase, in der kubische Probekörper mit der Kantenlänge 100mm bei unterschiedlichen Lagerungsbedingungen statisch und dynamisch gestaucht wurden. In nachfolgenden Versuchen wird der Einfluß der Probengröße ermittelt sowie die Eignung der verwendeten Materialmodelle anhand realistischer Belastungssituationen validiert.
Versuchsprogramm.
Rubbers are widely used as main sealing materials for containers for low and intermediate level radioactive waste and as additional component to metal seals in spent fuel and high active waste containers. The save encapsulation of the radioactive container inventory has to be guaranteed according to legislation and appropriate guidelines for long term storage periods as well as down to temperatures of -40 °C during transport.
Therefore the understanding of failure mechanisms that lead to leakage at low temperatures is of high importance.
It is known that the material properties of rubbers are strongly temperature dependent. At low temperatures this is caused by the rubber-glass transition (abbr. glass transition). During continuous cooling the material changes from rubber-like entropy-elastic to stiff energy-elastic behaviour, that allows nearly no strain or retraction, due to the glass transition. Hence rubbers are normally used above their glass transition but the minimum working temperature limit is not defined precisely, what can cause problems during application. Therefore the lower operation temperature limit of rubber seals should be determined in dependence of the material properties.
The results of Differential Scanning Calorimetry (DSC) and Dynamic Mechanical Analysis (DMA) are combined with the results of standardized measurements as the compression set according to ISO 815. To reduce the test time of the standard tests a faster technique was developed.
Additionally, the breakdown temperature of the sealing function of complete O-ring seals is measured in a component test setup to compare it with the results of the other tests. The experimental setup is capable of measuring the leakage rate at low temperatures by the pressure rise method.
A fluorocarbon rubber (FKM) was selected for this investigation as it is often used for radioactive waste containers. Some materials (seals and test sheets) were purchased from a commercial seal producer and some materials were compounded and cured at BAM in form of rubber sheets.
The aging management system for the mechanical components of nuclear power plants (NPPs) must be established and used by the licensee in such a way that the quality of safety relevant components is guaranteed for the completely designed lifetime of the NPP. This demands an extensive plant life management with special emphases on the knowledge of the degradation in material properties. The basic safety concept in Germany observes this circumstance. Lifetime extension of the German NPPs is an aim of the current valid coalition agreement of the German government. Operational extension of interim storage facilities requires, in comparison to the aging management system for NPP, an aging management system adapted to the Special circumstances of spent fuel storage casks. Extension of interim storage periods for spent fuel casks beyond the designed lifetime requires, in comparison to the components of an NPP, an increasing knowledge of material degradation with potential impact on cask integrity, e.g. leak tightness. Dry interim storage in Germany has been approved for 40 years. After that time, according to the present strategy, a final repository should be available. However, until now, such a final facility still does not exist, and the German exploration and licensing process is heavily delayed. Currently, discussions are continuing regarding further exploration of the Gorleben salt mine. There is willingness to overcome this situation that is clearly described in the available coalition agreement of the federal government. Anyway, however, the prediction is viewed; a repository for heat generating radioactive waste in Germany will not be available in the near future and may not be available when first storage facilities and casks reach their 40 years of approved lifetime, which will occur in ~25 years starting from now. Therefore, the question must be asked:
what has to be done with the existing storage casks in the interim facilities? May these casks be fit for continued use, with an extension of the storage period? One option is to have an aging management system, which creates enough information about the technical condition of safety relevant cask properties. This is the basis for safety evaluation for extended storage periods. In the present paper, possible aging mechanisms for high level waste storage casks are discussed, as well as the influence of the time dependent changes of the component properties.
Within the last years BAM has carried out numerous drop tests with prototype casks made of ductile cast iron onto targets according to the requirements for final disposal of non-heat generating waste in the German KONRAD repository. The results have shown that the target specifications in the acceptance criteria have to be defined more accurately to get reproducible test results with high precision. Hence, a suitable test stand foundation was developed with much effort. The integrity of the upper concrete layer of this target must be preserved during a test.
Recently the geometrical properties of a tested cubic cast iron container led to a concentration of the impact forces beneath the container walls. The target was damaged strongly with the consequence of inadmissible reduction of cask stresses. For that reason the target construction was modified. However, the basic design was not changed. A prefabricated concrete slab was still joined by a mortar layer to the IAEA target of the BAM drop test facility. In the course of the optimization of the test stand foundation the concrete slab dimensions and the reinforcement were enlarged. During the drop test repetition the target kept intact. Additionally, the mechanical behavior of the cast iron container and the target was analyzed by finite element calculations.
This improved target construction is suggested as a reference target for drop tests with casks whose mass and base area are covered by the container types VI or VII respectively according to the KONRAD repository acceptance criteria.
The measurements during the drop tests with cast iron casks have provided the strains on the cask surface at selected positions. This allows the verification of finite element simulations of drop tests which show the stress distribution also inside the component. In September 2008 a drop test was carried out with a cylindrical cast iron cask containing an artificial material defect which was designed under consideration of critical stress states in the cask body. This drop test could demonstrate the safety against failure by fracture of a cask made of a special cast iron with reduced fracture toughness.
Mechanical drop test scenarios for Type B (U) packages according to the IAEA regulations have to be carried out onto the so-called “unyielding target” (usually with cask impact limiters) and onto the puncture bar respectively. They are predefined and do not require any further investigation of scenarios that really could happen on transportation routes. Cask accident scenarios in the framework of approval procedures for interim storage sites are derived from a detailed analysis of the handling procedures necessary from arrival of cask at the site to its storing position.
In that case, casks are usually handled without impact limiters. Dependent on possible drop heights, drop positions and floor properties, conservative cask accident scenarios are derived for further safety proofs.
According to the mechanical assessment concept of the considered approval procedure numerical calculations have to be provided by the applicant to demonstrate mechanical cask safety. Stresses and strains in the cask body as well as in the lid System have to be identified and assessed. Using the example of a 3-mvertical-drop of a transport and storage cask for spent fuel elements onto the floor construction made of damping concrete covered by screed, BAM developed a finite element model. The finite element code ABAQUS/Explicit™ was used. Results of experimental investigations are not available. Therefore parameter studies are necessary to identify the sensitivity of the finite element model to significant Parameters and to verify the finite element models according to the requirements of the Guidelines for the Numerical Safety Analyses for the Approval of Transport and Storage Casks for Radioactive Materials (BAM GGR-008).
The paper describes the modeling of the material behavior and attachment of bottom side cask components. Questions concerning the modeling of a crack length limiting reinforcement in the screed layer are discussed. The influence of the mesh density of the screed layer and its strength is considered as well. Finally, the developed finite element model can be used for a numerical safety assessment. It can help to understand the complex mechanisms of the interaction between the cask components and floor construction.
The deformation and damage behaviour of damping concrete under impact loading conditions is investigated experimentally and numerically. The material model is based on the assumption of crushable foam with volumetric or isotropic hardening combined with ductile and shear damage criteria. Model parameters are determined in static and dynamic compression tests of confined cubic specimens. The derived material model is validated by numerical simulation of penetration tests. The static and dynamic penetration of indenters into uniform as well as assembled bricks made of damping concrete is discussed. Finally, the successful calculation of a large-scale drop test with a heavy cask-like test object onto a realistic damping concrete foundation is demonstrated.
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.
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.
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.
The numerical Simulation of transport and storage casks for radioactive waste in accidental scenarios requires adequate material models and input parameters for all relevant structural cask elements, the Impact foundation as well as possibly further components involved in the incident. They form the basis for reliable analysis of different loading situations allowing substituting, at least in parts, experimental investigations that are needed in safety Reports submitted for nuclear licensing procedures.
Hereby, the efficiency of the energy absorbing components, such as cask Impact limiters or equivalent structures placed onto or into the foundation of storage buildings, determine to a large extend
the performance and thus the design of the casks. The limiters, which are usually made of wood, polyurethane foam (PUF) or damping concrete, are subjected to Major plastic deformations, whose Distribution and magnitude strongly depend on loading speed, course and temperature.
Consequently, it is necessary to establish a systematic data basis and implement this information in advanced numerical material formulations in Order to enable finite element simulations to account for all relevant factors.
Finite element analysis (FEA) has been carried out for investigation of damping concrete under different impact loading conditions with a built-in material model and damage criteria available in FEA code ABAQUS.
At first, all parameters for the selected material model had been derived from compression Tests of cubic specimens. After that, a validation was carried out with different static and dynamic penetration tests. Finally, a 5 meter real drop test with a 23 Mg cylindrical cask could successfully be simulated.
Up to the end of this decade, corresponding to the planned date of starting operation in the final disposal KONRAD for non-heat generating waste in Germany, a lot of efforts are needed to condition and package the radioactive waste in containers certified by BfS (Federal Institute for Radiation Protection). This waste is produced by public sector and industry as well as nuclear energy facilities, which result in more than a half of the actually declared quantity growing especially after the phase out decision of nuclear power production in Germany and the subsequent decommissioning of nuclear power plants. BAM (Federal Institute for Materials Research and Testing) acts as responsible authority on behalf of BfS for design testing under consideration of the KONRAD requirements. Within the assessment procedure of containers, BAM has to verify the application documents, including material qualification, container geometry, corrosion protection, leakage rate and operational and accidental loading, and to evaluate quality assurance measures. Besides the previous completed approvals for various containers such as steel sheet and cast iron box-shaped containers and concrete cylindrical ones, BAM is actually carrying assessments for other types such as cast iron cylindrical containers and “old” steel sheet box shaped ones. The so called “old” containers present already loaded containers without any KONRAD certification, currently stored at licensed interim storages. In the benefit of the container assessment, BAM operates design testing facilities for drop and fire tests which are also used for research objectives to improve and expand the evaluation methods such as research project ConDrop. The latter deals with numerical simulations and analyses tools for further precise predictions about unfavorable drop test scenarios, safety margins, and design sensitivities for steel sheet containers. Furthermore, during the assessment procedure, several specifications about the containers, the inventory to be disposed and the required safety level have been intensely debated by authorities and applicants. Based on its experience in qualifying containers, BAM has been commissioned to identify insufficiently specified aspects in the waste acceptance criteria and to propose clearer definitions with regard to the secondary regulations in the planning approval notice. This paper focuses on aspects of the contribution of BAM in certifying KONRAD waste containers. This will include the current state of design testing procedures, evaluation methods that are available or are being developed and the main topics within the KONRAD requirements being actually specified by BAM.
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. 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. However, 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 behaviour 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 aluminium) 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 behaviour for additional test parameters are explained.
Overwiev about BAM
(2014)
A Memorandum of Understanding (MOU) between the Federal Institute for Materials Research and Testing (BAM) and Argonne National Laboratory (Argonne) was signed in the fall of 2014. Its objectives are to promote cooperation among scientists and specialists at BAM and Argonne and establish a framework for collaboration in the field of advanced technologies associated with the back end of the nuclear fuel cycle. Collaborative activities involving the BAM and Argonne may be implemented through the promotion of joint research activities and scientific workshops and conferences; the exchange of technical information; and visits by scientists, specialists, and graduate, postgraduate, and Ph.D. students. The first Argonne/BAM workshop was held at Argonne in October 2014 to discuss aging management issues related to the long-term dry storage of spent fuel. Major topics for potential collaboration on the extended storage of spent nuclear fuel and its subsequent transportation were identified.
Elastomer seals are widely used as barrier seals in containers for low and intermediate level radioactive waste and for spent fuel transportation casks. In addition, they are also used for spent fuel storage and transportation casks (dual purpose casks (DPC)) as auxiliary seals to allow leakage rate measurements of metal barrier seals for demonstration of their proper assembling conditions. Depending on the area of use, the rubber materials have to demonstrate proper sealing performance with regard to mechanical, thermal, and environmental conditions as well as irradiation during the entire operation period. Concerning DPC, degradation effects should be limited in a way that, for example, effects from potentially released decomposition elements may not harm e.g. metal barrier seals. Leakage rate measurements should be possible also after long interim storage periods prior to subsequent transportation.
Because of the complex requirements resulting from the various applications of containers for radioactive waste and spent nuclear fuel, BAM has initiated several test programmes for investigating the behaviour of elastomer seals. In this contribution the current status is described and first results are discussed.