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Eingeladener Vortrag
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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)
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.
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.
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.
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)
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.
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.
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.
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.
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.
The German repository “Konrad” is approved for the final storage of approximately 300.000 m³ of non-heat generating radioactive waste considering also the expected residues from operation and decommissioning of current power plants. According to the latest survey as constituted end of 2014, a large proportion of this volume already exists, of which 100.288 m³ are fully loaded packages. As prerequisite for their “Konrad” acceptance, the vessels have to be designed against operational and accidental conditions of the site and be manufactured according to adequate quality assurance programs. The respective safety assessment performed by BAM on behalf of the Federal Office for Radiation Protection is a major challenge especially for those casks that had been produced and loaded already some decades ago. Typically, they were manufactured in big numbers over a longer period of time involving a lot of constructional and management changes. This fact in combination with the previously rather low documentation standards often causes relevant uncertainties about the specific physical and geometrical characteristics of the cask materials and components as well as about details concerning their whole assemblage.
This paper highlights current assessment strategies for such older vessels including cubic steel sheet containers as well as cylindrical thick-walled casks. As compliance with each “Konrad”-requirement can hardly be shown independently, only integrated approaches promises success: First, a broader concept of data collection has to be implemented that allows closing documentation gaps by linking dispersed and diverse data, e.g. delivery slips and handwritten fabrication records, by including plausibility considerations as well as random inspection programs. Thereby, the relevant properties of all applied casks have to be identified and their potential range should be quantified as accurately as possible. This knowledge is necessary to define one or more representative configurations of the cask as the base for design testing. The actual investigation program must enable to track the influence of deviations and determine safety margins especially in the likely case that not the most conservative setting of all cask features can be tested. The paper evaluates the respective ongoing efforts of industrial applicants and summarizes BAM experiences and associated research projects.
In Germany, for the transport and storage of spent nuclear fuel, casks with double closure lid systems are used, which are equipped with Helicoflex® metal seals. The original interim storage period for these casks was planned to be 40 years. However, recent political developments indicate that a storage time of more than 80 years might be necessary. Therefore, the current storage licenses have to be renewed in due course, which requires extended knowledge of the long-term behaviour of all cask components.
At the Bundesanstalt für Materialforschung und -prüfung (BAM), metal seals have been investigated after ageing at temperatures between room temperature and 150 °C for up to 7 years. At regular intervals the seals have been tested for leakage rate as well as for their mechanical behaviour. From these tests pronounced influence of ageing could be shown which results in the decrease of the seal force and the useable resilience. The main reason attributed to these performance changes is the increasing permanent deformation of the outer layer of the seal due to creep. So far, an explicit analytical description of the long-term behaviour that could be used for predictions exceeding the tested ageing times and temperatures was not suitable.
In addition to the component tests that were already conducted, a comprehensive investigation programme has been started to describe the mechanical behaviour of the individual seal components.
The individual seal components are tested regarding different material characteristics including creep, deformation and microstructural properties. The comparability of the material characterization is ensured by comparing the basic raw material of the specimen with the actual component material.
Furthermore, detailed investigations of the time and temperature dependent deformation behaviour are realized by ageing seal segments at multiple temperatures for different periods of time and employing standardized tests.
By analysing the acquired data in combination with the ongoing long-term component tests a dataset for analytical prediction of the long-term behaviour of the seals is intended for an extended period of time under a certain temperature regime.
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.
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.
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.
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.
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&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.
This Presentation adresses the major needs and challenges of qualification processes for nuclear waste packages performed by BAM. Packages include those for high level radioactive waste like spent fuel as well as those for low and intermediate level waste to be disposed of in the Konrad repository.
Beginning with the current German nuclear waste management policy design testing capabilities and experience at BAM is explained followed by illustrating the major strategic process to identify future challenges including specific R&D needs.
Der Vortrag illustriert den Stand und die Herausforderungen im Rahmen der Behälter-Bauartprüfungen für das Endlager Konrad. Ausgehend von den Grundlagen der Bauartprüfung und die aktuelle Situation bei den Antragsverfahren werden die fachlichen Begutachtungsschwerpunkte sowie Maßnahmen zur Optimierung Prozess- und Verfahrenssteuerung erläutert.