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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.
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.
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.
Elastomers are widely used as the main sealing materials for containers for low and intermediate level radioactive waste and as an additional component to metal seals in spent fuel and high active waste containers. The safe encapsulation of the radioactive container inventory has to be guaranteed according to regulation and appropriate guidelines for long term storage periods as well as for temperatures as low as -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 elastomers are strongly temperature dependent. At low temperatures, this is caused by the rubber-glass transition (abbreviated: glass transition). During continuous cooling, the material changes from a rubber-like entropy elastic to a stiff energy elastic behaviour, which allows nearly no strain or retraction. Hence, rubbers are normally used above their glass transition, but the minimum working temperature limit is not defined precisely; this can cause problems during the above noted applications. Therefore, the lower operation temperature limit of elastomer seals must be determined in dependence of the material properties. Differential scanning calorimetry (DSC) and dynamic mechanical analysis (DMA) are combined with standardised measurements to determine the compression set according to ISO 815. To reduce the test time of the standard tests, a faster technique than normally used was developed. Additionally, the breakdown temperature of the sealing function of complete O ring seals was measured in a component test set-up to compare it with the results of the other tests. The experimental set-up 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. The materials (seals and test sheets) were purchased from a commercial seal producer.
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.
Sicherheitsbewertung von Lagerbehältern bei Auslaufen einer befristeten Zwischenlagergenehmigung
(2013)
Bei der sicherheitstechnischen Begutachtung der behälterspezifischen Aspekte nach Ablauf einer Zwischenlagergenehmigung ist entsprechend einer Neubeantragung zu verfahren. Die Sicherheitsbewertung erfolgt nach dem aktuellen Stand von Wissenschaft und Technik bei Berücksichtigung der zur Behälterbauart bereits vorliegenden Sicherheitsnachweise und Begutachtungsergebnisse sowie der physikalischen und chemischen Alterungseinflüsse der bisherigen Zwischenlagerung. Am Beispiel einer beantragten weiteren Zwischenlagerung nach Auslaufen einer auf 20 Jahre befristeten Genehmigung und dem notwendigen Behältertransport zu einem neuen Zwischenlager werden Erfahrungen bzgl. der durchgeführten Sicherheitsbewertung zur Erfüllung der Technischen Annahmebedingungen des neuen Behälterlagers dargestellt.
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.
In Germany the concept of dry interim storage of spent fuel and vitrified high active waste in dual purpose metal casks is implemented (currently for periods of up to 40 years). The casks being used have an approved package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the guidelines of the German Reactor Safety Commission for dry interim storage of spent fuel. Currently part of the assessment process of the cask design for transport on public routes is to evaluate the suitability of the used materials with respect to their properties and their compatibility for possible transport periods. For transport on public routes during or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfill the specifications of the transport approval or other sufficient properties which satisfy the proofs for the compliance of the safety objectives at that time. This paper describes the state-of-the-art technology in Germany and points out arising prospective challenges and which related questions have to be answered in future. Present research and knowledge concerning the long term behavior of transport and storage cask components (such as gaskets and shielding components) have to be consulted as well as experiences from interim cask storage operations. Challenges in the safety assessment are therefore i.e. the behavior of aged metal and elastomeric gaskets under IAEA test conditions to ensure that the results of drop tests can be transferred to the compliance of the safety objectives at the time of transport after the interim storage period (aged package). Assessment methods for the material compatibility, the behavior of fuel assemblies and the aging behavior of shielding parts (e. g. influence of radiation) are further issues as well.
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.
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.
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.
Due to delays in the site-selection procedure to establish a deep geological repository for spent nuclear fuel and high level waste and in construction of the already licensed Konrad-repository for low and intermediate level waste without heat generation, 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 caslcs or as primary seals for low and intermediate level waste packages is an important issue in this context (Jaunich, 2013; Jaunich, 2014; Kömmling, 2015). The polymeric structure of these seals causes a complex mechanical behaviour with time-dependent elasticity reduction and loss of elastic recovery.
The paper presents first results of a comprehensive test Programme consisting of several static and dynamic mechanical short- and long-term tests which have been carried out at BAM on specimens made of representative types of elastomers, fluorocarbon rubber (FKM) and ethylene propylene diene rubber (EPDM). The investigation of the test results used to identify material models and their Parameters as well as the development of two finite element models for the numerical Simulation of tension and compression tests using the finite element code ABAQUS® are described. The calculation results are presented in comparison to the test results. The influence of important material and test parameters was investigated and discussed in sensitivity analyses.
It is important to evaluate the effect of thermal ageing on the sealing performance of metal gaskets under high temperature for long-term usage. Therefore, in order to gain representative data for this kind of metal-sealed lid System, BAM is currently performing laboratory tests of different gasket types with aluminum and silver jackets at three different temperatures under static conditions up to four years so far, using test flanges for gaskets with full scale cross section diameter but much smaller outer diameter. On the other hand, in order to investigate the applicability of the numerical methodology to evaluate the long-term behavior of the metal gaskets, such as a correlation between seal pressure force and holding time, CRIEPI is developing a modeling method including material tests (tensile and creep tests) at high temperature. In this paper, the applicability of the finite element method (ABAQUS) to predict the recovery displacement and residual seal pressure force of the gasket complex was verified by comparing the calculated values with BAM’s laboratory test results under the joint research agreement between BAM and CRIEPI.
Existing spent nuclear fuel (SF) and high active waste (HAW) management policies and practices worldwide are the result of past presumptions that sufficient reprocessing and/or disposal capacity would be available in the near term. Consequently, in the past many countries have developed specific solutions for different periods of time due to their individual national nuclear policies.
In Germany the concept of dry interim storage in dual purpose metal casks before disposal is being pursued for SF and HAW management and transport and storage licenses have been issued accordingly. The current operation licenses for existing storage facilities have been granted for a storage period of up to 40 years. This concept has demonstrated its suitability for over 20 years so far. Relevant safety requirements have been assessed for the short-term as well as for the long-term for site-specific operational and accidental storage conditions. But in the meantime significant delays in the national repository siting procedure occurred which will make extended storage periods necessary in the future.
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.
Im hier vorliegenden Bericht wurde im Rahmen einer umfassenden Literaturrecherche der nationale und internationale Stand von Wissenschaft und Technik hinsichtlich Behälterkonzepten und Behälteranforderungen erfasst. Dabei wurde eine umfangreiche Dokumentensammlung zusammengestellt und sortiert, die als Quellenbasis für alle weiteren Arbeiten zur Verfügung steht. Die Auswertung der gesammelten Informationen resultiert in einem Überblick über internationale Behälterentwicklungen sowie einer detaillierten Übersicht über die Behälterkonzepte und die ihnen zugrundeliegenden Anforderungskataloge aus elf fortgeschrittenen Endlagerprogrammen.
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.
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. But in the meantime longer storage periods might be necessary for the future and therefore additional material data will be required. BAM is involved in the qualification and evaluation procedures of those seals from the early beginning. Because long-term tests are always time consuming BAM has early decided to perform additional tests with specific test seal configurations to gain a better understanding of the long-term behavior with regard to seal pressure force, leakage rate and useable resilience which is safety relevant mainly in case of accidental mechanical loads inside a storage facility or during a subsequent transport. Main test parameters are the material of the outer seal jacket (silver or aluminum) and the temperature. This paper presents the BAM test program including an innovative test mock-up and most recent test results. Based on these data extrapolation models to extended time periods are discussed, and also future plans to continue tests and to investigate seal behavior for additional test parameters are explained.
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.
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.
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.
Langzeitversuche über 10 Jahre an Federkern-Metalldichtungen mit Aluminium- bzw. Silberummantelungen
(2023)
Federunterstützte Metalldichtungen mit Aluminium(Al)- oder Silber(Ag)-Ummantelung werden u.a. in Behältern für Wärme entwickelnde radioaktive Abfälle eingesetzt, da diese Dichtungen eine sehr gute Dichtheit gewährleisten sowie Langlebigkeit und Beständigkeit gegenüber erhöhten Temperaturen und radioaktiver Strahlung aufweisen. Auch wenn die Sicherheit solcher Behälter und der verwendeten Dichtungen vielfach belegt wurde, sind sie Gegenstand andauernder Forschung, etwa im Hinblick auf eine absehbar benötigte verlängerte Zwischenlagerdauer. Aus diesem Grund werden an der Bundesanstalt für Materialforschung und -prüfung (BAM) im Fachbereich 3.4 „Sicherheit von Lagerbehältern“ seit über 20 Jahren Versuche an solchen Metalldichtungen durchgeführt. Dabei sollen zusätzliche Erkenntnisse hinsichtlich der Sicherheitsreserven der Dichtungen in unterstellten Störfallszenarien (axiale Bewegung des Deckelsystems bzw. Aufweitung der Nutgeometrie) und insbesondere detailliertere Erkenntnisse zum Langzeitverhalten gewonnen werden.
Langzeituntersuchungen an Metalldichtungen mit innen liegendem Federkern und doppelter Ummantelung
(2013)
In Deutschland werden in Transport- und Lagerbehältern (TLB) wärmeentwickelnde radioaktive Abfälle gegenwärtig über einen Zeitraum von bis zu 40 Jahren zwischengelagert. Aufgrund der Situation der noch nicht abgeschlossenen Endlagersuche werden hier möglicherweise Verlängerungen in Betracht zu ziehen sein. Im Deckeldichtsystem der TLB werden in der Regel Metalldichtungen vom Typ Helicoflex® eines französischen Herstellers verwendet, die aus einem innen liegenden Spiralfederkern und zwei Ummantelungen bestehen. Die Ermittlung der Fliess- und Relaxationseigenschaften der Dichtungskomponenten über den vorgesehenen langen Betriebseitraum sind vor dem Hintergrund des Erhalte eines geforderten hohen Dichtniveaus ein wichtiger Untersuchungsgegenstand.
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.
Metal seals are in common use as an essential component for the leak tight closure of transport and storage casks for spent fuel and vitrified high active waste from reprocessing. They are placed between the massive monolithic cask body made of ductile cast iron or forged steel and the primary and secondary lids, which are fixed with screws around the lid circumference, in order to ensure long term safe enclosure of the radioactive cask inventory. Hereby, long term reliable pressure forces in combination with a tight surface contact are as important as well maintained seal and surface qualities and dry and clean conditions during seal assembly and operation. A few years ago BAM started systematic investigation on the long term seal behavior with re-spect to longer storage periods and seal behavior in accidental cask scenarios with dynamic de-formations of the seal groove geometry or short term drop in pressure force up to lid lifting. BAM has developed test flanges for seals with full scale cross section diameter but much small-er outer diameter to gain representative data for metal-sealed lid systems. They are placed in appropriate testing machines for relevant mechanical loading under static and dynamic condi-tions with the ability to measure the standard helium leakage rate continuously during each test phase. This paper focuses on BAM long term tests of different seal types with aluminum and silver jackets at three different temperatures of +20°C, +100°C and +150°C under static conditions over longer periods of time. Due to creeping effects, a reduction of the pressure force does ap-pear during loading and unloading depending on prior holding times. So far, test results of up to 2 years indicate a clear correlation between seal pressure forces and holding time, temperature and seal type allowing for extrapolating to much longer periods of time. The paper presents basic correlations, test and evaluation procedures as well as important provisional results and gives an outlook on further investigations.
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.
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.
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. According to appropriate guidelines and regulations safe enclosure of the radioactive container contents has to be guaranteed for long storage periods as well as down to temperatures of -40 °C for transportation. Therefore the understanding of seal behavior is of high importance.
In this paper we focus on the behavior of elastomer seals at low temperatures with regard to poten-tial decrease of leak-tightness. In addition, changes in material properties due to aging effects over long periods of time and their influence on the seal performance is investigated.
It is known that material properties of rubbers are strongly temperature dependent. At low tempera-tures this is caused by the rubber-glass transition (abbr. glass transition). During continuous cool-ing, the material changes from rubber-like entropy-elastic to stiff energy-elastic behavior, that al-lows 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.
Aging of elastomer seals is important, as possible dynamic loads may have to be considered during the whole interim storage period (so far approved in Germany for up to 40 years) and for transporta-tion after storage.
For the investigations, fluorocarbon (FKM) and ethylene-propylene-diene (EPDM) rubbers were selected as they are often used in radioactive waste containers. Some materials were purchased from a commercial seal producer and some materials were compounded and cured at BAM. The elastom-ers where studied by several thermo-analytical methods and compression set to characterize the material behavior at low temperatures. Additionally component tests were performed to determine the breakdown temperature of the sealing function of complete elastomer O-rings.
Der Bericht zum Arbeitspaket 3 baut auf den Ergebnissen der internationalen Recherche (AP1) sowie der Zusammenstellung der Randbedingungen und Beanspruchungsgrößen für die Endlagerbehälter (AP2) auf. Die Anforderungen an die Behälter – regulatorische, betriebliche sowie solche aus der Standortgeologie und den Einwirkungen und Prozessen, denen die Behälter ausgesetzt sind – werden hergeleitet und systematisch dargestellt.
Die für das Endlagersystem zu betrachtenden Nutzungsphasen werden aus den regulatorischen Anforderungen hergeleitet und charakterisiert. Die Einwirkungen auf die Behälter in den drei potenziellen Wirtsgesteinen sowie die sich für die Erfüllung der Sicherheitsfunktionen des Behälters als Teil des Endlagersystems ergebenden Behälterfunktionen werden ebenfalls wirtsgesteins- und zeitphasenabhängig betrachtet. Schließlich werden aus den Einwirkungen auf die Behälter und den Anforderungen an die Behälterfunktionen auch quantifizierbare Behälterfunktionen für alle drei potenziellen Wirtsgesteine abgeleitet.
German approach and feedback on experience of transportability of SNF packages after interim storage
(2014)
In Germany, the concept of dry interim storage of spent fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years. The casks being used have an approved package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the recently revised 'Guidelines for dry interim storage of irradiated fuel assemblies and heat-generating radioactive waste in casks' by the German Waste management Commission. For transport on public routes between or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfil the specifications of the transport approval or other sufficient properties, which satisfy the proofs for the compliance of the safety objectives at that time. In recent years, the validation period of transport approval certificates for manufactured, loaded and stored packages were discussed among authorities and applicants. A case dependent system of 3, 5 and 10 years was established. There are consequences for the safety cases in the Package Design Safety Report, including evaluation of long term behaviour of components and specific operating procedures of the package. The present research and knowledge concerning the long term behaviour of transport and storage cask components have to be consulted as well as experiences from interim cask storage operations. Challenges in the safety assessment are e.g. the behaviour of aged metal and elastomeric gaskets under IAEA test conditions to ensure that the results of drop tests can be transferred to the compliance of the safety objectives at the time of transport after the interim storage period. Assessment methods for the material compatibility, the behaviour of fuel assemblies and the aging behaviour of shielding parts are issues as well. This paper describes the state of the art technology in Germany, explains recent experience on transport preparation after interim storage and points out arising prospective challenges.
German approach and experience feedback of transport ability of SNF packages after interim storage
(2013)
In Germany the concept of dry interim storage of spent fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years.
The casks being used have an approved package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the recently revised “Guidelines for dry interim storage of irradiated fuel assemblies and heat-generating radioactive waste in casks” by the German Waste management Commission (ESK).
For transport on public routes between or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfil the specifications of the transport approval or other sufficient properties which satisfy the proofs for the compliance of the safety objectives at that time.
In recent years the validation period of transport approval certificates for manufactured, loaded and stored packages were discussed among authorities and applicants. A case dependant system of 3, 5 and 10 years was established. There are consequences for the safety cases in the Package Design Safety Report including evaluation of long term behavior of components and specific operating procedures of the package.
Present research and knowledge concerning the long term behavior of transport and storage cask components have to be consulted as well as experiences from interim cask storage operations. Challenges in the safety assessment are e.g. the behavior of aged metal and elastomeric gaskets under IAEA test conditions to ensure that the results of drop tests can be transferred to the compliance of the safety objectives at the time of transport after the interim storage period.
Existing spent nuclear fuel (SF) and high active waste (HAW) management policies and practices worldwide are the result of past presumptions that sufficient reprocessing and/or disposal capacity would be available in the near term. Consequently, in the past many countries have developed specific solutions for different periods of time due to their individual national nuclear policies.
In Germany the concept of dry interim storage in dual purpose metal casks before disposal is being pursued for SF and HAW management and transport and storage licenses have been issued accordingly. The current operation licenses for existing storage facilities have been granted for a storage period of up to 40 years. This concept has demonstrated its suitability for over 20 years so far. Relevant safety requirements haven been assessed for the short-term as well as for the long-term for site-specific operational and accidental storage conditions. But in the meantime significant delays in the national repository siting procedure occured which will make extended storage periods necessary in the future.
This paper describes the current situation in Germany with regard to dry cask storage and focuses on current perspectives considering regulatory, technical, and scientific aspects for storage license renewal. Since there is one case of a storage license limitation to only 20 years due to administrative reasons, first experience is currently gathered in case of an expiring storage license. Subsequent license options have been pursued intensively including the extension of the initial storage license as well as shipping all casks to another storage facility considering an extended storage period at that storage facility as well. All safety relevant aspects have to be reviewed on basis of the current state-of-the-art which might be different from the initial safety demonstrations. That includes new safety assessment standards as well as improved knowledge base. Major issues are e.g. improved accident scenario analyses, assessment methods, and consideration of aging effects from previous operation periods. Inspection programs with respect, e.g. to bolts and seals to verify leak-tightness and lid screw pre-stress have been initiated to demonstrate proper cask conditions for extended storage as well as transportation to another storage facility.
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.
Dieser Bericht untersucht die im ersten Arbeitspaket ermittelten vorhandenen bzw. entwickelten Behälterkonzepte dahingehend, inwieweit die jeweiligen Behälterfunktionen unter den im zweiten und dritten Arbeitspaket ermittelten Randbedingungen und Einwirkungen den im dritten Arbeitspaket ermittelten Anforderungen an endlagerfähige Behälter in einem deutschen Endlager genügen. Zum anderen sollen, basierend auf den im zweiten Arbeitspaket ermittelten Randbedingungen und den im dritten Arbeitspaket hergeleiteten und zusammengestellten Anforderungen an die Endlagerbehälter, Vorschläge für mögliche generische Endlagerbehälter in den drei potenziellen Wirtsgesteinen Steinsalz, Tonstein und Kristallingestein erarbeitet werden. Die im ersten Arbeitspaket recherchierten Behälterkonzepte, deren Entwicklungslinien und zugrundeliegende Behälteranforderungen bilden hier den Ausgangspunkt für weitergehende Überlegungen hinsichtlich ihrer potenziellen Eignung für ein zukünftiges deutsches Endlager und zur Skizzierung generischer Konzepte für zukünftige Endlagerbehälter.
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.
In Germany dual purpose casks for spent nuclear fuel (SF) or high level radioactive waste (HLW) are used for safe transportation and interim storage. Key safety issues in both fields are the safe enclosure of the radioactive material, the safe removal of decay heat, securing nuclear criticality safety, limitation of radiation exposure to acceptable levels and keeping it as low as reasonable achievable (ALARA principle).
Whereas these safety requirements during transportation are ensured by the Type-B(U)-design testing and approval procedure on basis of internationally agreed IAEA (International Atomic Energy Agency) requirements storage licenses are issued on national and site specific safety analyses.
This paper presents experiences from recent interim storage safety evaluation and licensing procedures in Germany on basis of a Type-B(U) certified cask designs concerning specific differences from Operation and accident conditions inside the storage facility. The focus is laid on the interaction of cask and building structures with regard to shielding, heat removal and accident analyses including aircraft crash. Basic safety is assured by thick-walled metal casks with monitored double lid Systems. They also reduce radiation to levels where workers can safely operate and maintain the casks inside the storage facility. The storage building e. g. provides additional shielding and protection against extemal hazards depending on the building construction. In addition, the storage building helps to reduce radiation levels at the boundary of the storage site well below regulatory limits given by the German Radiation Protection Ordinance (StrSchV).
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.
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.
Component and material investigations on metal seals for high level radioactive waste containers
(2018)
In Germany casks for interim storage of spent nuclear fuel and high level radioactive waste are equipped with Helicoflex® metal seals as main sealing barrier of the double lid system. The long-term behaviour of those seals is investigated at Bundesanstalt für Materialforschung und –prüfung (BAM) in order to evaluate the safety function of the containment at different temperatures over storage periods of 40 years or more in case extended interim storage becomes necessary.
Long-term investigations have been done for ageing times of up to 8.5 years at temperatures ranging from room temperature to 150 °C. It was found that the seal force and useable resilience decrease over time, which is mainly caused by creep deformation of the aluminium (or silver) outer jacket of the seals. This effect becomes stronger with increasing temperature.
The ageing processes of the seal material and the overall seal behaviour is under investigation in order to derive analytical descriptions for the long-term seal performance. Thus, standardized tests on the basic seal materials, with focus on aluminium, and additional investigations on the seals as a component are conducted. The current investigations include compression and tension creep measurements as well as tensile testing. Furthermore, ageing of seal segments provides information on the development of the contact area width, jacket thickness and microstructural changes in dependence of time and temperature. The obtained data are used for the development of material models and an analytical approach to describe and predict the time and temperature dependent sealing behaviour in the long-term.
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.
Die Bundesanstalt für Materialforschung und –prüfung (BAM) wurde von der AG 3 der Kommission Lagerung hoch radioaktiver Abfallstoffe gebeten, Anforderungen an Behälter für die Endlagerung hoch radioaktiver Abfallstoffe im Hinblick auf ih-re Rückholbarkeit während des Endlagerbetriebes und ihre Bergbarkeit nach Ver-schluss des Endlagers über einen Zeitraum von 500 (bis 1.000 ?) Jahren darzu-stellen. Die BAM kommt diesem Wunsch auf der Grundlage ihrer langjährigen Er-fahrungen und Fachkenntnisse auf den Gebieten der Bauartprüfung und der si-cherheitstechnischen Begutachtung von Transport- und Lagerbehältern für radio-aktive Stoffe nach.
Bei der Auswahl eines Standortes für ein Endlager für hochradioaktive Abfälle (Wärme entwickelnde radioaktive Abfälle und ausgediente Brennelemente) in Deutschland werden auf Grundlage des Standortauswahlgesetzes von 2017 die drei potenziellen Wirtsgesteine Steinsalz, Tonstein und Kristallingestein in Betracht gezogen. Die Auswahl eines konkreten Endlagerstandortes steht dabei in engem Zusammenhang mit den für die jeweiligen Wirtsgesteine geeignet erscheinender Endlagersystemen. Zentraler Bestandteil dieser Endlagersysteme sind die Endlagerbehälter, deren Gestaltung und Auslegung maßgeblich die Randbedingungen für die Transport- und Einlagerungstechnik vorgibt und zudem die Nachweise zur Betriebs- und Langzeitsicherheit entscheidend prägt. Daher ist es für die Behälterentwicklung erforderlich, die an Endlagerbehälter zu stellenden Anforderungen systematisch herzuleiten. Auf dieser Grundlage können anforderungsgerechte Endlagerbehälter zielgerichtet für die jeweils zu untersuchenden Endlagerstandorte in unterschiedlichen Wirtsgesteinen entwickelt werden.
Dieser Beitrag fasst die wesentlichen Ergebnisse des von BAM und BGE TECHNOLOGY GmbH im Zeitraum 2017 bis 2020 bearbeiteten und vom Bundesministerium für Wirtschaft und Energie (BMWi) geförderten Forschungsvorhabens KoBrA zusammen, die in einem Abschlussbericht und vier zugehörigen Teilberichten umfassend dokumentiert sind.
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.
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.