Filtern
Erscheinungsjahr
Dokumenttyp
- Beitrag zu einem Tagungsband (110)
- Vortrag (88)
- Zeitschriftenartikel (26)
- Forschungsbericht (7)
- Beitrag zu einem Sammelband (6)
- Posterpräsentation (3)
- Buchkapitel (2)
- Monografie (1)
- Sonstiges (1)
Sprache
- Englisch (135)
- Deutsch (108)
- Mehrsprachig (1)
Schlagworte
- Behälter (18)
- Endlager (16)
- Safety (13)
- Interim storage (11)
- Metal seal (11)
- Storage (11)
- Bauartprüfung (10)
- Simulation (9)
- Spent fuel (9)
- Disposal (8)
Organisationseinheit der BAM
Eingeladener Vortrag
- nein (88)
Component and material investigations on metal seals for high level radioactive waste containers
(2018)
Long-term investigations performed at BAM look to extend the state of knowledge on safety-related components of interim storage containers. Metal seals act as the primary sealing barrier in the bolted double lid closure system of the containers. The behaviour of metal seals has been investigated for ageing times up to 8.5 years and for various temperatures. The main cause for reduction in useable resilience overtime was due to creep deformation of the outer jacket of the seal.
The investigation of the long-term performance of sealing systems employed in containers for radioactive waste and spent nuclear fuel is one research focus area for division 3.4 “Safety of Storage Containers” at the Bundesanstalt für Materialforschung und -prüfung (BAM). Our investigations comprise investigations on metallic and elastomeric seals and covers experimental investigations to get a database on the component/material behaviour, work on analytical descriptions and numerical analysis. Our aim is to understand the long-term behaviour of the sealing systems for evaluation of their performance during possible extended interim storage and subsequent transportation.
This presentation starts with a brief overview about the current situation of spent nuclear fuel and high level waste storage in Germany followed by a discussion of the expected timeline towards final disposal. Furthermore an overview is presented about research programmes for radioactive waste disposal and also extended interim storage issues due to the delay of the national repository siting procedure. In the second part of the presentation BAM research activities related to extended interim storage are addressed. This includes the status and preliminary results of proactive R&D projects concerning metal seals, degradation effects of polymers used as neutron shielding materials and finally elastomer seals.
This presentation addresses safety aspects concerning dual purpose casks for transportation and storage of spent fuel and high level radioactive waste. The long term performance of casks and their safety relevant components like sealed lid systems and the long term performance of cask internals, especially spent fuel assemblies are discussed. Transportation after interim storage is another key issue and due to the fact of delayed disposal projects current interim storage periods need to be extended causing additional safety demonstration needs for longer periods of time. Regarding this situation present BAM research activities in this area and their preliminary outcomes are presented.
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.
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.
Bisherige Betriebserfahrungen und Erkenntnisse lassen die grundsätzliche Eignung der Transport- und Lagerbehälter auch über 40 Jahre hinaus erwarten.
Eine Verlängerung der genehmigten Zwischenlagerdauer erfordert eine Neugenehmigung. Hierbei sind die zu diesem Zeitpunkt sicherheitstechnisch relevanten Eigenschaften der tatsächlich vorhandenen Inventare und Behälter zu berücksichtigen.
Sicherheitstechnische Nachweise sind auf Grundlage belastbarer Daten und Erkenntnisse zu führen. Diese müssen rechtzeitig verfügbar sein (z.B. Betriebserfahrungen oder zusätzliche F&E-Programme z.B. zu Behälterdichtsystemen oder Inventarverhalten).
Alterungsmanagementprogramme und periodische Sicherheits-überprüfungen während der Zwischenlagerung haben sowohl für die verlängerte Zwischenlagerung als auch für nachfolgende Transporte eine erhebliche Bedeutung.Die Schaffung eines spezifischen Regelwerks für die verlängerte Zwischenlagerung von Transport- und Lagerbehältern und deren nachfolgendem Abtransport unter Berücksichtigung der Anforderungen aus beiden Verwendungsbereichen erscheint sinnvoll.
Eine verlängerte Zwischenlagerung dient ausschließlich der Überbrückung bis zur Endlagerung. Hierbei sollten die bestehenden Behälter und Inventare frühzeitig in die Überlegungen zur Definition von Anforderungen für die Endlagerung einbezogen werden.
Die internationalen Entwicklungen und Untersuchungen im Bereich der längerfristigen Zwischenlagerung sollten zur Nutzung von Synergieeffekten weiter verfolgt, ausgewertet und mitgestaltet werden.
Im Zusammenhang mit dem Ausstiegsbeschluss der Bundesregierung zur Nutzung
der Kernenergie und der zugehörigen Vereinbarung mit den Energieversorgungsunternehmen
vom 11. Juni 2001 wurde auch die Beschränkung der Entsorgung auf die
direkte Endlagerung bestrahlter Brennelemente sowie die Verpflichtung zur Errichtung
und Nutzung von dezentralen Zwischenlagern an den Kraftwerksstandorten
vereinbart. Dieser Verpflichtung wurde mit der Beantragung und Genehmigung von
insgesamt 12 Standortzwischenlagern bis Ende 2003 sowie vorlaufend 4 sog. Interimslagern
entsprochen. Darüber hinaus bestehen weiterhin die bereits genehmigten
zentralen Zwischenlager in Ahaus und Gorleben sowie das AVR-Behälterlager Jülich
und das Zwischenlager Nord (ZLN) in Lubmin bei Greifswald.
Die Zwischenlagerung bestrahlter Brennelemente und hochradioaktiver Abfälle aus
der Wiederaufarbeitung erfolgt in Deutschland in Transport- und Lagerbehältern.
Diese Behälter verfügen einerseits als Transportbehälter über eine verkehrsrechtliche
Zulassung des Bundesamtes für Strahlenschutz (BfS) auf der Grundlage einer
umfassenden Bauartprüfung durch die BAM gemäß den Vorgaben des Gefahrgutrechts
auf der Basis der Internationalen IAEA-Regulations. Andererseits ist die Zwischenlagerung
dieser Behälter in den o. g. Zwischenlagern auf der Grundlage des
Atomgesetzes (AtG) vom BfS gesondert genehmigt.
Voraussetzung für die Annahme und Einlagerung der beladenen Behälter in ein Zwischenlager
ist u.a. die Einhaltung zahlreicher technischer Anforderungen an die Qualität, die Eigenschaften und den Zustand der Behälter und deren Inventar. Diese
Anforderungen sind in den jeweiligen lagerspezifischen „Technischen Annahmebedingungen“
festgeschrieben. Sie reichen von den Nachweisen über die ordnungsgemäße
Herstellung jedes einzelnen Behälters einschließlich aller seiner Komponenten
bis hin zu den Abfertigungsvorschriften zur Montage der Deckel-Dichtsysteme nach
der Behälterbeladung im KKW.
The safety assessment of casks for radioactive material at interim storage facilities or in final repositories includes the investigation of possible handling accidents if clearly defined test conditions are not available from the regulations. Specific handling accidents usually are the drop of a cask onto the transport vehicle or the floor as well as the collision with the wall of the storage building or another cask. For such load cases an experimental demonstration of cask safety would be difficult. Therefore, numerical analyses of the entire load scenario are preferred. The lessons learnt from dynamic finite element analyses of accident scenarios with thick-walled cubical containers or cylindrical casks are presented. The dependency of calculation results on initial and boundary conditions, material models, and contact conditions is discussed. Parameter sets used should be verified by numerical simulation of experimentally investigated similar test scenarios. On the other hand, decisions have to be made whether a parameter or property is modeled in a realistic or conservative manner. For example, a very small variation of the initial impact angle of a container can cause significantly different stresses and strains. In sophisticated cases an investigation of simpler limit load scenarios could be advantageous instead of analyzing a very complicated load scenario.
Metal gaskets in the lid system of transport and storage casks for radioactive material have to guarantee leak-tightness and safe enclosure of the radioactive inventory under normal and accident conditions during transport, in case of accidents and for the longterm interim storage. For safety assessments by nurnerical simulation of the thermomechanical behaviour of a lid system with metal gaskets, the finite element method offers three options to use: i) special gasket elements, ii) complex three-dimensional modeHing with solid elements, and iii) a simplified axisymmetric approach. Gasket elements can be adjusted by many parameters, but they give only a global representation of the experimentally observed gasket behaviour. For calculations of the entire cask including the lid system with gaskets, nonetheless this approach is recommended. To investigate the hardly measurable impact behaviour in detail or to extrapolate the long-term behaviour, a local modeHing of all parts of a gasket and their interactions with effects like elastic-plastic deformation, creep, relaxation, and friction is necessary. The three-dimensional model can describe the change of contact area between outer jacket of the gasket and flange dependent on the load conditions, what overcomes an essential limitation of special gasket elements. To simplify the problern of investigating the underlying physical effects, an axisymmetric lid system can be modelled with axisymmetric finite elements. Usually an approximation is only necessary for the helical spring of a metal gasket. This paper explains basic ideas for an adequate finite element simulation of cask lid systems with metal gaskets and their thermo-mechanical behaviour under specific load scenarios.
Metal seals of O-ring form are often used in lid-systems of transport and storage casks for radioactive waste in Germany. To investigate their mechanical behaviour, three dimensional (3D) finite element (FE) models were created using solid elements for all of the seal components. The material behaviour of each component is described with a unified static elastic-plastic material model. The total strain is defined as the sum of linear elastic strain and plastic strain with power-law hardening. The model was carefully validated by comparison of Simulation results with experimental results. The influence of material fluctuation of each seal component due to varying properties and the sensitivity of different seal diameters on the seal force are analysed and discussed.
The results show that the material properties of helical spring have major influence on seal force and that the influence of seal diameters is negligible small in the studied range. This is very important to use the test results from seals with small diameter for the assessment of seal behaviour with larger diameter as used in transport and storage cask.
The safety evaluation of cask components made of ductile cast iron includes investigations to prevent brittle fracture. Generally, ductile cast iron is endangered by brittle fracture especially at low temperatures (down to -40°C) and in combination with existing crack-like material defects. An applicable method is the assessment of fracture resistance using fracture mechanics according to the IAEA guidelines. The approach is based on the prevention of fracture initiation. For application of these principles for drop loads, account must be taken both of dynamic stresses within the component and dynamic material behavior. Basically, the dynamic stress intensity factor of postulated pre-existing crack-like defects is compared with the dynamic fracture toughness of the material. Applicable numerical and experimental methods for the safety assessment of cask components are demonstrated for the case of an artificially pre-cracked cylindrical cast iron cask which undergoes dynamic loading conditions as result of the hard impact between the cask and a concrete target. The proposed evaluation procedure is a combination of numerical and experimental steps. Exemplarily, the calculated stress intensity factor is compared with measured fracture toughness values from single edge notched bending specimens.
The evaluation of cladding integrity is a major issue to be demonstrated in Germany for extended interim storage periods up to 100 years and subsequent transportation considering operational and accidental conditions with respect to reactor operation, cask drying and dry interim storage. The chemical reaction between the zirconium fuel cladding and the cooling water in water-cooled reactors produces hydrogen and zirconium oxide. Hydrogen diffuses into the cladding and precipitates as zirconium hydrides when the solubility limit is reached, preferably oriented in hoop direction. At high temperatures during vacuum drying procedures, the hydrides can dissolve. Over a succeeding period of slow cooling with existing hoop stress the hydrides precipitate again, but partly reoriented along the radial direction of the cladding. This change of microstructure in combination with a decreasing temperature (0.5...2 K/year) during (extended) interim storage and additional mechanical load by handling procedures or under accident conditions could lead to a potential cladding embrittlement and consequently increased failure probability. The current research project BRUZL (Fracture mechanical analysis of spent fuel claddings under long-term dry interim storage conditions) has been launched by BAM to investigate potential sudden brittle failure of spent fuel claddings at small deformation under long-term dry interim storage conditions and is presented.
In Germany, spent nuclear fuel and high active waste from reprocessing is stored in transport and storage containers with double lid systems that are equipped with metal seals completing the primary sealing barrier. The tasks of the Bundesanstalt für Materialforschung und -prüfung (BAM) within the interim storage licensing procedures ruled by the German Atomic Energy Act include the long-term safety evaluation of the container design regarding the permanently safe enclosure of the inventory. In order to generate a knowledge base for the safety evaluation, research regarding the long-term behavior of the critical components is performed. So far, the containers are licensed for an interim storage period of 40 years. However, due to significant delays in establishing a final repository, the required time span for interim storage is expected to increase significantly. Thus, a widespread investigation program is run to gain systematic data on the long-term behavior of the seals and to develop prediction models.
Long-term seal investigations consider the development of their restoring seal force, their useable resilience and their achievable leakage rate caused by aging at temperatures ranging from room temperature up to 150 °C. This year, the total time span of the tests reaches 10 years. Furthermore, seal segments are aged at the selected temperatures for up to 300 days. From these segments additional information on the sealing behavior, changes of the seal contact and the material behavior is gained.
This contribution deals with the current results of the long-term seal investigations at BAM. Furthermore, insights of the more in-depth component and material investigations of the metal seals with focus on the seal contact development are discussed and the ongoing work aiming for an analytical description of the thermo-mechanical aging effects on metal seals are presented.
This publication summarizes the work carried out during an IAEA coordinated research project (CRP) on safe and reliabe management of spent nuclear fuel. This work contributes to the overall goal of demonstrating the performance of spent nuclear fuel and related storage systems components over long durations, and facilitates the transfer of this knowledge to Member States. The technical areas addressed by the CRP participants were related to potential degradation mechanisms in metal casks and concrete overpacks, such as stress corrosion cracking; long term integrity and performance of the fuel cladding; thermo-mechanical behaviour of the metal seals; and long term gamma and neutron shielding capability. The participants also considered past and ongoing demonstration programmes. For each technical topic addressed the main conclusions have been drawn, including information on the development of specific monitoring and inspection techniques as well as future opportunities for closing relevant data gaps. A major achievement is also the establishment of a worldwide network of experts working on current research projects to demonstrate the long term performance of spent fuel in dry storage systems.
The investigation of the long-term performance of sealing systems employed in containers of radioactive waste and spent nuclear fuel is one research focus area for division 3.4 \Safety of Storage Containers" at the Bundesanstalt für Materialforschung und -prüfung. The behaviour of metallic seals, which are employed in interim storage casks for spent nuclear fuel and high active waste, over time is of high importance for ensuring safe enclosure. Therefore, investigations on these systems were started at BAM to get a general understanding of the relevant processes. Our investigations comprise investigations on different parameters which influence the seal Performance and the main part is focussed on the time and temperature dependent behaviour. Our aim is to understand the longterm behaviour of the sealing systems for evaluation of their performance during possible extended interim storage and subsequent transportation. In this contribution an overview of the performed tests and their respective results is given. A focus lies on a comparison between seals with large torus diameter in comparison to small diameters.
Safety of long term containment and storage of radioactive waste is an essential issue, which is increasingly gaining international attention. Extending interim storage beyond initial license periods is one of the major challenges worldwide.
BAM has been involved in all kinds of radioactive waste container safety evaluation for storage and transportation including their long term performance from the very beginning. Today, all work related to the safe management of radioactive waste is linked by the activity field “Nuclear Waste Management” within the Focus Area “Energy” at BAM and in conjunction with a research strategy and agenda. Most recently, the working group “Long Term Safety of Container Systems and Components” was established to launch and perform research projects related to safety issues concerning the extended interim storage of radioactive waste.
For example, within our project LaMEP, we investigate ageing effects of metal and elastomer seals as well as polymeric neutron shielding materials, which are container components that fulfil major safety functions. The results of the experiments help to understand the long term behaviour of these components. Likewise, our project BRUZL started at BAM in 2018 with the aim of developing a fracture mechanics approach to understand and describe a potential brittle failure mechanism of fuel cladding during long-term interim storage. BAM also contributes to the GRS project “Development of methods and models and international exchange for safety evaluation in the frame of interim storage of radioactive waste”. Additionally, BAM is involved in several international research programmes dealing with extended interim storage, such as the European Joint Programme on Radioactive Waste Management (EURAD), the Extended Storage Collaboration Program (ESCP) and IAEA Coordinated Research Projects.
The present contribution provides a comprehensive overview of BAM research activities contributing to national and international gap analyses and to filling knowledge gaps regarding the safe long term management of radioactive waste.
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.
Influence of impact angle and real target properties on drop test results of cubic containers
(2017)
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.
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 approach of storing spent nuclear fuel in dual purpose casks until final disposal is based on individually licensed storage sites concerning specific nuclear inventories and dual purpose cask types. The currently approved license period is maximum 40 years.
With nuclear policy changes including the phase-out decision of 2011 and a new approach for the HLW repository siting process set up 2013 by law also the spent fuel and HLW management strategy has to be adapted. This includes the timely development of disposal container concepts as part of the overall repository siting process and the need for longer interim storage periods until final disposal. Other aspects affect new regulations about the financing of the nuclear waste management by the industry and the public sector, the restructuring of responsible authorities in 2016/2017 and the adaption of R&D funding programs addressing also pre-disposal (or extended interim storage) issues.
BAM as a Federal Research Institute has been responsible for or in charge of container safety Evaluation for transport and storage casks for radioactive waste for several decades and started addressing potential Extended interim storage issues already several years ago. This included participation in international projects at the IAEA or in the U.S. Extended Storage Collaboration Program (ESCP) as well as long term laboratory tests with metal seals, elastomeric seals and polymers for neutron shielding purposes. Test results of BAM are published regularly and gained data provide valuable information about thermo-mechanical and radiological degradation effects as well as the long term Performance of respective cask components. Regarding the reliable extrapolation and prediction of the long term Performance of cask materials and components for time periods of up to 80 or even 100 years the continuation and extension of laboratory tests and the development of analytical and numerical simulation models and tools is essential and under way.
With the clear expectation of the need for extending spent fuel storage in Germany beyond the initial license period of maximum 40 years additional material data and knowledge about the long term performance and integrity of casks and internals like fuel rods will be required to meet major safety goals and regulatory requirements.
Based on a broad knowledge base about fuel cladding behavior, degradation mechanisms and related potential failure mechanisms, e. g. induced by reoriented Hydride precipitation, as illustrated in this paper, a specific gap analyses for the German situation regarding fuel conditions and cladding materials should identify the specific knowledge gaps to be addressed and closed in the near future. This is of specific interest as all spent fuel assemblies are going to be transferred to dry cask storage with the next decade and as a consequence of the German nuclear phase-out decision.
Within several decades, BAM as a Federal materials research institute gathered a broad range of expertise in the safety evaluation of materials, components and casks under operational and accidental thermal and mechanical conditions. This paper elucidates selected Basic considerations regarding technical and scientific knowledge about the long term aging performance and potential failure mechanisms of fuel cladding materials.
Under consideration of the national R&D funding policy which has been adapted to the extended interim storage or pre-disposal needs and the competences of BAM, its future R&D strategy is explained and some R&D objectives have already been identified. BAM will continue and expand ist engagement in the area of investigating fuel cladding material behavior in the long term and developing validated simulation and analyses tools in collaboration with national and international partners. R&D outcomes shall improve the knowledge-base to close identified knowledge gaps in the area of demonstrating fuel cladding integrity and to provide crucial information with regard to future safety assessments for the extended interim spent fuel storage.
Low and intermediate level radio active waste has been produced in Germany for several decades. While a part of the waste had been deposited in the Morleben repository and the Asse mine which have been operated for just a limited period of time, the major amount is conditioned and packaged in temporarily stored waste containers.
Those containers are designated for final disposal in the KONRAD repository being currently under construction and expected to start operation in 2022. As a consequence, an assessment of the waste container documentation and safety is strongly needed to receive an approval for the KONRAD repository. This requalification is challenging because those old containers have to meet the same safety requirements as new containers.
This paper describes and discusses the major challenges of spent fuel management in Germany after the phase-out of nuclear electricity generation was decided in 2011 and a new repository siting procedure was implemented in 2013. Consequences from those decisions which were legally founded by amendments of the German Atomic Energy Act (AtG) result in the need to transfer all remaining spent fuel from limited reactor operation (last reactor shutdown until the end of 2022) into casks for subsequent dry interim storage on-site. Storage licenses are generally issued site-specific considering specific dual purpose casks (DPC) and their inventories and they are generally limited to 40 years so far. But the need for extending the interim storage period in the future has become obvious. Even though, this may not be an issue to be solved already today questions about additional safety demonstrations will arise as soon as licenses need to be extended. Certainly, these questions will ask for reliable data about the long term performance and safety of structures, systems, and components, e. g. the long term performance of cask components and materials like bolted closure systems including metal seals, or fuel rod behaviour concerning cladding materials under stress and temperature conditions. In case of dual purpose casks for storage and transportation this includes aspects on how to demonstrate transportability during or after several decades of interim storage. Long term investigations often require plenty of time and therefore need to be initiated timely.
This paper describes and discusses the major challenges of spent fuel management in Germany after the phase-out of nuclear electricity generation was decided in 2011 and a new repository siting procedure was implemented in 2013. Consequences from those decisions which were legally founded by amendments of the German Atomic Energy Act (AtG) result in the need to transfer all remaining spent fuel from limited reactor operation (last reactor shutdown until the end of 2022) into casks for subsequent dry interim storage on-site. Storage licenses are generally issued site-specific considering specific dual purpose casks (DPC) and their inventories and they are generally limited to 40 years so far. But the need for extending the interim storage period in the future has become obvious. Even though, this may not be an issue to be solved already today questions about additional safety demonstrations will arise as soon as licenses need to be extended. Certainly, these questions will ask for reliable data about the long term performance and safety of structures, systems, and components, e. g. the long term performance of cask components and materials like bolted closure systems including metal seals, or fuel rod behaviour concerning cladding materials under stress and temperature conditions. In case of dual purpose casks for storage and transportation this includes aspects on how to demonstrate transportability during or after several decades of interim storage. Long term investigations often require plenty of time and therefore need to be initiated timely.
The safe and secure long term storage of spent nuclear fuel until its final disposal in a deep geological repository is a corner stone of the German nuclear waste management strategy. So far interim storage licenses are limited to 40 years concerning the initial concept of having a repository available until 2035. But in the meantime the exploration of the designated Gorleben salt dome was finally stopped and a reset of the complete site selection process was concluded by the German Federal parliament and subsequently fixed by law. A 33-member “Commission on the storage of highly radioactive materials” was established in 2014, to elaborate a consent based siting process and to develop criteria for the safe disposal of heat generating waste. The disposal commission agreed and published its final report in 2016.
As a major consequence, the necessary timeframe for establishing the final repository is going to be exceeded significantly and with that, an extension of the current dry interim storage in dual purpose casks becomes necessary. This paper explains the major actions to be taken in Germany to address all relevant issues for a future extension of storage licenses beyond the initial timeframe of 40 years.
Perspectives on extending interim storage of spent nuclear fuel (beyond the initial license period)
(2017)
The presentation adddresses the consequences and challenges of the German nuclear phase-out in conjunction with the restart of siting process for a high-level radioactive waste repository. Concerning spent fuel this will lead to extended interim storage periods for the dry storage casks beyond the initial licensing period of 40 years.
For extending storage periods additional safety assessments and data about the long term performance of materials, components and casks will be needed and BAM has already been lauching specific R&D projects addressing identified data gaps.
Elastomer seals are widely used as barrier seals in containers for low and intermediate level radioactive waste and in some countries for spent fuel transportation casks. 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.
In Germany the low and intermediate level radioactive waste is planned to be disposed in the Konrad repository in standardized containers. Many of these containers use an elastomer seal between lid and container body. Depending on the radiactive inventory and a possible containment of the waste in inner barrels and/or concrete structures, such containers may require a specific leak tightness. In such cases, specified leakage rates are shall not be exceeded by the the seals to keep potential activity releases from the inventories well below defined limits.
Generally, all waste containers have to fulfill the waste acceptance criteria for the Konrad repository as defined by the reports BfS SE-IB-29/08 and SE-IB-30/08. From these criteria several requirements for seal performance are derived. The seals have to:
- ensure full functionality until final container disposal,
- maintain its function in case of a thermal or mechanical accident,
- ensure encapsulation also at -20 °C.
This combination of requirements is challenging especially as a seal exchange, which is typical for many other applications, is hardly possible and thus a sound understanding of how to perform safety evaluation of containers is required. Therefore BAM has initiated investigations on elastomer seal behaviour. This work focuses on low temperature behaviour and sealing performance concerning static conditions as well as partial loss of compression and ageing of elastomer seals. Concerning ageing, property changes caused by long-term thermooxidative conditions as well as gamma irradiation are investigated.
This contribution presents an overview and explanation of the aims of the ongoing research and gives an update of the most relevant results.
Der Vortrag vermittelt einen Überblick über internationale Forschungsaktivitäten im Hinblick auf zukünftig deutlich zu verlängernde Zwischenlagerzeiträume für bestrahlte Brennelemente. Ausgehend von der aktuellen Entsorgungssituation werden die relevanten Fragestellungen skizziert. Davon ausgehend werden die FuE-Aktivitäten der IAEA vorgestellt sowie die wichtigsten Themen beim Extended Storage Collaboration Program (ESCP) in den USA. Schließlich wird das "European Joint Programme on Radioactive Waste Management and Disposal“ kurz vorgestellt und werden die nationalen Initiativen auf diesem Sektor diskutiert.
Elastomers show a high versatility which makes them ideal materials for sealing applications in various fields. Especially under changing application conditions the high recovery potential of this class of material is beneficial to compensate temperature or pressure fluctuation, and geometrical changes resulting from mechanical loads in e.g. accident conditions. Out of these reasons elastomers are also used in containers for low and intermediate level radioactive waste and for spent fuel transportation casks. In casks designed for low and intermediate level waste elastomer seals can act as primary seal responsible for the containment function whereas in spent fuel storage and transportation casks (dual purpose casks (DPC)) elastomer seals are used as auxiliary seals to allow leakage rate measurements of metal barrier seals. An inherent prerequisite for this kind of application is the Long time-scale of operation without or with limited possibility of seal replacement. In Germany an interim storage license for DPC`s is typically issued for 40 years, a timeframe which might increase in the future due to challenges of the final repository siting procedure. For low and intermediate level waste, also long time periods are required before final disposal can be achieved. Therefore, the performance of elastomer seals over extended time periods is, as for other applications, of high importance.
A typical approach to ensure long-term functionality is to perform accelerated aging tests to calculate an estimated lifetime by assuming e.g. Arrhenius like equations for the timetemperature relationship. This approach requires a suitable end of life criterion considering the application of interest. This often can represent a challenge on its own.
As BAM is involved in most of the cask licensing procedures and especially responsible for the evaluation of cask-related long-term safety issues we initiated several test programs for investigating the behavior of elastomer seals. Experiments concerning the low temperature performance down to -40 °C and the influence of gamma irradiation have been started first.
Currently the thermal aging behavior of elastomer seals, which is the topic of this contribution, is examined. For our aging investigations we use a broad approach to first determine the property changes in different elastomer materials due to thermo-oxidative aging at elevated temperatures and secondly, we test how the typical methods of lifetime extrapolation can be applied to these results. This Approach enables us to detect and exclude undesired side effects which very often influence lifetime estimations. In this contribution, our recent results are extended. The results show that lifetime estimation based on single material properties can be misleading and therefore a combination of several methods is recommended.
In Germany spent nuclear fuel is stored and transported in casks that possess bolted lid systems equipped with double jacket metal seals of Helicoflex® type in order to enable safe enclosure of the radioactive inventory. A dry interim storage period of 40 years was planned and the casks are licensed for that time span. However, due to political reasons and delayed disposal projects this time span is expected to increase significantly. It is therefore necessary to evaluate the long-term sealing behavior of the casks exceeding 40 years of operating time. In this paper, we discuss our approach to investigate the aging behavior of such metal seals, and to predict the long-term sealing behavior.
Accelerated aging component tests with seals are performed at temperatures up to 150°C to investigate the course of seal force, useable resilience and leakage rate. Even though these tests have already shown a significant decrease in seal force and useable resilience after aging times of less than 1 year, for most seals the leakage rate stayed better than the specified value. However, observation of the leakage rate has to be continued to determine the aging effect on the seal properties. As a main influence on the seal long-term behavior the increasing permanent deformation of the outer jacket has been recognized. It is made of aluminum or silver which means that the temperatures that are relevant for the aforementioned application in casks lead to creep deformation.
For further investigation of seal behavior a comprehensive research program concerning the material characteristics and aging behavior of the individual seal components, with an emphasis on the outer jacket material, has been launched. The aim is to get a better understanding of the time and temperature dependent creep mechanisms and deformation. The examinations are separated into tests on seal segments and tests on standardized specimen.
In order to be able to study the aging process, seal segments are compressed in flanges and stored at temperatures ranging from room temperature to 150°C. After defined time intervals the segments are analyzed by using standardized tests including structure investigation and creep tests. The results are compared with the behavior of the specimen basic raw material. This is done to ensure comparability of the thin jacket material and additional sheet material used for standardized tests. Thus, the prediction of component changes after different aging times at different temperatures can be made. As a result forecasts on the long term seal behavior are intended.
Due to delays in the siting procedure to establish a deep geological repository for spent nuclear fuel and high level radioactive waste as well as in construction of the already licensed Konrad repository for low and intermediate level radioactive waste, extended periods of interim storage become more relevant in Germany. BAM is involved in most of the cask licensing procedures and especially responsible for the evaluation of cask-related long-term safety issues.
The long-term performance of elastomer seals for lid systems of transport and storage casks, whether used as auxiliary seals in spent fuel casks or as primary seals for low and intermediate level waste packages, is an important issue in this context. The polymeric structure of these seals causes a complex mechanical behavior with time-dependent sealing force reduction.
The results of a comprehensive purpose-designed test program consisting of basic compression and tension tests as well as relaxation tests on unaged specimens of representative types of elastomers (fluorocarbon rubber (FKM) and ethylene propylene diene rubber (EPDM)) at different temperatures and strain rates are presented. They were used to identify the constitutive behavior and to obtain parameters for finite element material models provided by the computer code ABAQUS®. After estimating the influence of uncertainties such as Poisson’s ratio and friction coefficient by sensitivity analyses, the chosen parameters had to prove their suitability for the finite element simulation of the specimen tests themselves.
Based on this preliminary work the simulation of a specific laboratory test configuration containing a typical elastomer seal with circular cross section is presented. The chosen finite element material model and the implemented parameters had to show that they are able to represent not only the specimen behavior under predominantly uniaxial load but also the more complex stress states in real components. Deviations between the measured and calculated results are pointed out and discussed.
For the consideration of long-term effects in the simulation of elastomer behavior, test results of aged specimens are needed. First information about a new test program, started recently and planned to provide these data, are given.
The Bundesanstalt für Materialforschung und –Prüfung (BAM) runs an investigation program on the long-term behavior of multi-component metal seals. Such seals are used in a wide area of applications including transport and storage casks for spent nuclear fuel and high level radioactive waste.
The seal function is mainly based on the compression of the inner helical spring, which generates the necessary seal force to keep the sealing surfaces in close contact. This in turn leads to a plastic deformation of the outer jacket of the seal, comprised of highly ductile aluminum or silver that adapts to the sealing surfaces of cask body and bolted lid, thus providing high Level leak tightness. In Germany, those casks are licensed for Interim storage periods of up to 40 years or more if extended Interim storage would become necessary before a final repository is available. Thus, the sealing performance has to be evaluated, including factors like elevated temperature due to decay heat or mechanical loads due to transport under normal as well as accident conditions. Long-term investigations at BAM have been running over the last nine years to identify and evaluate the seal performance by measuring the remaining seal force, the useable resilience and the leakage rate after various time intervals at temperatures ranging from room temperature up
to 150 °C. It was found that the seal force and useable resilience decrease with time and temperature, caused by creep deformation of the outer jacket. In order to obtain an analytical description for the seal behavior and to achieve more information on the material behavior under application conditions a comprehensive investigation program with Focus on aluminum as outer jacket material was launched. The program includes material investigations such as compression and tension creep tests with representative basic materials. An additional test setup allows for the continuous measurement of the remaining seal force at temperatures of up to 150 °C. Furthermore, seal segments are compressed and stored in heating chambers, thus producing segments at different stages of the aging process. The segments are investigated regarding the development of the contact area width, jacket thickness and microstructural changes. This data will be used to develop material models and an analytical description of the time and temperature dependent long-term sealing behavior. This paper explains the current status of gained test results and modelling approaches and closes with an outlook to the future Project plans.
Die Verfügbarkeit von Behältern für die Entsorgung nicht Wärme entwickelnder radioaktiver Abfälle im Endlager Konrad stellt ein wesentliches Element des nationalen Entsorgungskonzeptes für diese Abfallart dar. Grundlage für die Verfügbarkeit geeigneter Behälter sind typspezifische Bauartprüfungen, auf deren Grundlage eine Eignungs-feststellung in Form eines Prüfzeugnisses durch die Bundesgesellschaft für Endlagerung mbH (BGE), vormals Bundesamt für Strahlenschutz (BfS), erfolgt. Die Prüfgrundlagen sind in den durch das BfS (heute BGE) veröffent-lichten „Anforderungen an endzulagernde radioaktive Abfälle“ (Endlagerungsbedingungen) sowie der zugehörigen „Produktkontrolle radioaktiver Abfälle, radiologische Aspekte“ (PK) verankert. Die Eignungsfeststellung für eine Be-hälterbauart ist gleichzeitig die Grundlage für eine Behälterserienfertigung entsprechend den im Prüfzeugnis veran-kerten Spezifikationen und Randbedingungen.
Mit den Gesetzen zur Neuordnung der Organisationsstruktur im Bereich der Endlagerung (Mitte 2016) und zur Neuordnung der Verantwortung in der kerntechnischen Entsorgung (Mitte 2017) mit der Finanzierung des Kern-energieausstiegs wurden wesentliche Weichenstellungen im Hinblick auf den anstehenden Rückbau der bereits stillgelegten und in den kommenden Jahren noch stillzulegenden deutschen Kernkraftwerke vorgenommen. Bezüg-lich der nicht Wärme entwickelnden radioaktiven Abfälle führt dies zu einer spürbaren Intensivierung der Anstren-gungen zur Bereitstellung qualifizierter Gebinde für das Endlager Konrad, das nach aktueller Planung bis 2027 sei-nen Einlagerungsbetrieb aufnehmen soll. Dies wiederum führt zu einer erheblichen Verdichtung der Verfahren zur Behälterbauartprüfung, die allen Seiten (Antragsteller, BGE, Gutachter) erhebliche Anstrengungen abfordert.
Die Bundesanstalt für Materialprüfung und –forschung (BAM) ist auf Grundlage des Kap. 7 der o. g. „Produktkon-trolle …“ (PK) von der BGE mit den Behälterbauartprüfungen für das Endlager Konrad beauftragt. Hierbei ist eine Vielzahl von Anforderungen zu berücksichtigen, wozu neben den Anforderungen an Abfallbehälter gemäß Kap. 5 der Endlagerungsbedingungen weitere Anforderungen an die Verpackung (Kap. 3.2.2 PK), Kenngrößen (Kap. 3.5.1 PK) und schließlich die eigentliche Prüfung von Behältern/Verpackungen zählt (Kap. 7 PK). Letzteres beinhaltet die umfassende Behälterbeschreibung und –spezifikation, die Behälterauslegung, die thermischen und mechanischen Baumusterprüfungen (z.B. Brand- und Fallprüfungen), und schließlich die qualitätssichernden Maßnahmen für Her-stellung und Betrieb der Behälter. Neben der klassischen Bauartprüfung neuer Behältertypen spielt auch die Prü-fung bereits hergestellter Behälter, sog. Altbehälter, eine wesentliche Rolle, da sich etliche tausend Gebinde bereits in der Zwischenlagerung befinden und nach Möglichkeit für das Endlager Konrad qualifiziert werden sollen.
Dieser Beitrag berichtet aktuell über den Umfang und Stand der laufenden Bauartprüfungen bei der BAM unter Be-rücksichtigung der besonderen Herausforderungen und der zuletzt erzielten Fortschritte. Dabei erfordert die paral-lele Bearbeitung der zahlreichen Verfahren einen verstärkten Koordinierungsbedarf hinsichtlich vergleichbarer fach-licher Aspekte sowie hinsichtlich Priorisierung und Ressourceneinsatz. Bei den experimentellen Nachweisen (Fall- und Brandversuche) betrifft das hinsichtlich der Nutzung der Versuchseinrichtungen bei der BAM die Koordination mit anderen Prüfaufträgen, wie z. B. verkehrsrechtliche Baumusterprüfungen. Schließlich erfordern diverse Fach-fragen zur Interpretation der Anforderungen aus Endlagerungsbedingungen und Produktkontrolle intensive Abstim-mungsprozesse zwischen BAM, BGE und Antragstellern. Wichtige Ergebnisse mit übergeordneter Bedeutung wer-den in Form erläuternder Fachnotizen durch die BGE veröffentlicht und können so bei zukünftigen Antragsverfahren berücksichtigt werden.
Die Verfügbarkeit von Behältern für die Entsorgung nicht Wärme entwickelnder radioaktiver Abfälle im Endlager Konrad stellt ein wesentliches Element des nationalen Entsorgungskonzeptes für diese Abfallart dar. Grundlage für die Verfügbarkeit geeigneter Behälter sind typspezifische Bauartprüfungen, auf deren Grundlage eine Eignungs-feststellung in Form eines Prüfzeugnisses durch die Bundesgesellschaft für Endlagerung mbH (BGE), vormals Bundesamt für Strahlenschutz (BfS), erfolgt. Die Prüfgrundlagen sind in den durch das BfS (heute BGE) veröffent-lichten „Anforderungen an endzulagernde radioaktive Abfälle“ (Endlagerungsbedingungen) sowie der zugehörigen „Produktkontrolle radioaktiver Abfälle, radiologische Aspekte“ (PK) verankert. Die Eignungsfeststellung für eine Behälterbauart ist gleichzeitig die Grundlage für eine Behälterserienfertigung entsprechend den im Prüfzeugnis verankerten Spezifikationen und Randbedingungen.
Mit den Gesetzen zur Neuordnung der Organisationsstruktur im Bereich der Endlagerung (Mitte 2016) und zur Neuordnung der Verantwortung in der kerntechnischen Entsorgung (Mitte 2017) mit der Finanzierung des Kern-energieausstiegs wurden wesentliche Weichenstellungen im Hinblick auf den anstehenden Rückbau der bereits stillgelegten und in den kommenden Jahren noch stillzulegenden deutschen Kernkraftwerke vorgenommen. Bezüg-lich der nicht Wärme entwickelnden radioaktiven Abfälle führt dies zu einer spürbaren Intensivierung der Anstren-gungen zur Bereitstellung qualifizierter Gebinde für das Endlager Konrad, das nach aktueller Planung bis 2027 sei-nen Einlagerungsbetrieb aufnehmen soll. Dies wiederum führt zu einer erheblichen Verdichtung der Verfahren zur Behälterbauartprüfung, die allen Seiten (Antragsteller, BGE, Gutachter) erhebliche Anstrengungen abfordert.
Die Bundesanstalt für Materialprüfung und –forschung (BAM) ist auf Grundlage des Kap. 7 der o. g. „Produktkon-trolle …“ (PK) von der BGE mit den Behälterbauartprüfungen für das Endlager Konrad beauftragt. Hierbei ist eine Vielzahl von Anforderungen zu berücksichtigen, wozu neben den Anforderungen an Abfallbehälter gemäß Kap. 5 der Endlagerungsbedingungen weitere Anforderungen an die Verpackung (Kap. 3.2.2 PK), Kenngrößen (Kap. 3.5.1 PK) und schließlich die eigentliche Prüfung von Behältern/Verpackungen zählt (Kap. 7 PK). Letzteres beinhaltet die umfassende Behälterbeschreibung und –spezifikation, die Behälterauslegung, die thermischen und mechanischen Baumusterprüfungen (z.B. Brand- und Fallprüfungen), und schließlich die qualitätssichernden Maßnahmen für Her-stellung und Betrieb der Behälter. Neben der klassischen Bauartprüfung neuer Behältertypen spielt auch die Prü-fung bereits hergestellter Behälter, sog. Altbehälter, eine wesentliche Rolle, da sich etliche tausend Gebinde be-reits in der Zwischenlagerung befinden und nach Möglichkeit für das Endlager Konrad qualifiziert werden sollen.
Dieser Beitrag berichtet aktuell über den Umfang und Stand der laufenden Bauartprüfungen bei der BAM unter Be-rücksichtigung der besonderen Herausforderungen und der zuletzt erzielten Fortschritte. Dabei erfordert die paral-lele Bearbeitung der zahlreichen Verfahren einen verstärkten Koordinierungsbedarf hinsichtlich vergleichbarer fach-licher Aspekte sowie hinsichtlich Priorisierung und Ressourceneinsatz. Bei den experimentellen Nachweisen (Fall- und Brandversuche) betrifft das hinsichtlich der Nutzung der Versuchseinrichtungen bei der BAM die Koordination mit anderen Prüfaufträgen, wie z. B. verkehrsrechtliche Baumusterprüfungen. Schließlich erfordern diverse Fach-fragen zur Interpretation der Anforderungen aus Endlagerungsbedingungen und Produktkontrolle intensive Abstim-mungsprozesse zwischen BAM, BGE und Antragstellern. Wichtige Ergebnisse mit übergeordneter Bedeutung wer-den in Form erläuternder Fachnotizen durch die BGE veröffentlicht und können so bei zukünftigen Antragsverfahren berücksichtigt werden.
For the long-term storage of spent nuclear fuel and other high-Level radioactive waste, dual purpose casks for the transportation and storage are widely used. Usually, these casks consist of a thick-walled monolithic cask body and a bolted double barrier lid System equipped with metal seals (consisting of an inner helical spring and two outer metal layers with a total cross-section diameter between 5 and 10 mm) to ensure the long-term safe confinement of the radioactive inventory.
Based on analytical and experimental studies, such metallic Systems have been qualified of storage periods of up to four decades so far as long as proper manufacturing and assembling conditions are met. More than 20 years of Operation experience with a permanently increasing number of casks has shown full compliance with the expectations.
In the meantime many countries have to face a significant delay of establishing a high-level radioactive waste repository what leads to Need for extending interim storage periods for probably 60, 80 or 100 years.
For that reason the long-term performance of metal seals for such time periods is of specific internst to demonstrate their sufficient safety function and specific investigations have already been launched, e.g. by BAM, to gain more experimental data and to develop validated models to extrapolate the seal performance to longer periods of time and or requested temperature levels.
Suitable containers for receiving high-level radioactive waste are a key elements of every repository system concerning different host rocks like rock salt, claystone and crystalline rock as being considered in Germany. The disposal container design is decisive regarding boundary conditions for the transport, emplacement techniques, and the operational and long-term safety assessment. The systematic derivation of the requirements to be placed on disposal containers forms the basis for a targeted, comprehensible and transparent development of containers that meet the requirements. The presentation summarizes the outcomes of the research project KoBrA.
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.
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.
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.
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.
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.