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BAM as competent authority regarding assessment of mechanical and thermal design, activity release and quality assurance aspects of SNF and HLW transport packages developed a thesaurus of experience feedback topics from recent approval procedures. The list is structured according to the European PDSR guide. It involves issues, which from BAM point of view, needed clarification during last package design assessment procedures. The list contains issues from operation (e.g. deactivation of handling lugs not intended for package handling), maintenance (e.g. leak tightness of impact limiting devices) to technical assessment (e.g. formation of hydrogen by radiolysis and its impact on pressure and ignitability, consideration of ageing mechanisms, thermo-mechanical assessment, impact of gaps between content and flask, spent fuel behaviour) as well as to general and specific safety analysis report requirements (e.g. report structure and required data).
Three drop test campaigns have been performed with DN 30 Protective Structural Packaging (PSP) developed by DAHER NUCLEAR TECHNOLOGIES GmbH for the transport of natural, enriched and reprocessed uranium hexafluoride (up to 5 wt%) in 30B cylinders.
The mechanical prototype testing is intended to demonstrate that the package DN30 complies with regulatory requirements under normal and hypothetical accident conditions of transport (NCT, ACT) relevant to IF, AF and B(U)F packages, respectively.
The paper includes the results of the latest test campaign carried out in 2015-2016 at the drop test facility of BAM, Germany with new full scale prototypes of the DN30 PSP and 30B cylinders. Repetition of drop test sequences became necessary after changing to Polyisocyanurate foam as shock absorbing material with variable foam densities. Furthermore, the mechanical behavior of the UF6 content of the 30B cylinder is now simulated by a mixture of cement and steel grid as modified surrogate material; instead of small steel balls. The behavior of this new content simulation is assumed to be more realistic with respect to the properties of real UF6.
The long time-frame of the necessary safe enclosure of radioactive waste in containers gains interna-tionally an increasing attention. As well extended storage periods as a usually necessary subsequent transportation after storage have to be evaluated.
In Germany, the concept of dry interim storage in dual purpose metal casks before disposal is being pursued for spent nuclear fuel (SNF) and high level waste (HLW) management. The initially planned and established dry interim storage license duration of up to 40 years will be too short and its exten-sion will become necessary since there is no repository available in due time.
The main safety relevant components are the thick-walled dual purpose metal casks. These casks consist e.g. of a monolithic cask body with integrated neutron shielding components (polymers, e.g. polyethylene) and a monitored double lid barrier system equipped with metal and elastomer seals. The metal seals of this bolted closure system guarantee the required leak-tightness whereas the elastomer seals allow for leakage rate measurement of the metal seals.
At BAM we focus on the long-term performance of the sealing system and neutron shielding materials as their application conditions have to be concerned as rather demanding.
Therefore, BAM began to investigate the overall function and long-term behavior of elastomer and metal seals and also the irradiation influence on neutron shielding materials.
Recently a new project was initiated to expand our investigations to get more and in some context better applicable data.
This paper describes the approach of our new research project with the aim of further closure of existing knowledge gaps and presents an overview of investigations concerning the low temperature behavior of elastomer seals, of running long-term tests on elastomer seals at different temperatures under static conditions and the influence of gamma irradiation on neutron shielding materials.
Storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage. Beside consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case. Not only the technical objects are subject of aging but also the safety demonstration basis is subject of “aging” due to possible changes of regulations, standards and scientific/technical knowledge. The basic document, defining the transport safety conditions, is the package design safety report (PDSR) for the transport version of the DPC. To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the PDSR, in connection with periodic renewals of package design approval certificates, have to be carried out. The main reviewing tool is a gap analysis. A gap analysis for a PDSR is the assessment of the state of technical knowledge, standards and regulations regarding safety functions of structures, systems and components.
The design of a special form radioactive material has to resist a severe transport accident without undue loss or dispersal of radioactive material.
Safety assessment by authorities competent for design approval has to include besides the required test program (impact, percussion, bending and heat test) also the evaluation of the quality management system for design, manufacture, testing, documentation, use, maintenance and inspection.
These quality assurance measures have to assure that every specimen of the approved design is produced in the same verified quality and every specimen must be able to survive the severe mechanical and thermal tests without undue loss or dispersal of radioactive material at any time of its working life.
All important aspects in the design approval procedure by BAM as the competent authority for approvals of special form radioactive material in Germany are summarized in a guideline published in 2014. This paper will give additional explanations to some aspects in safety assessment, e.g.:
the applicability of leak test methods and the need to consider ageing aspects.
The German repository “Konrad” is approved for the final storage of approximately 300.000 m³ of non-heat generating radioactive waste considering also the expected residues from operation and decommissioning of current power plants. According to the latest survey as constituted end of 2014, a large proportion of this volume already exists, of which 100.288 m³ are fully loaded packages. As prerequisite for their “Konrad” acceptance, the vessels have to be designed against operational and accidental conditions of the site and be manufactured according to adequate quality assurance programs. The respective safety assessment performed by BAM on behalf of the Federal Office for Radiation Protection is a major challenge especially for those casks that had been produced and loaded already some decades ago. Typically, they were manufactured in big numbers over a longer period of time involving a lot of constructional and management changes. This fact in combination with the previously rather low documentation standards often causes relevant uncertainties about the specific physical and geometrical characteristics of the cask materials and components as well as about details concerning their whole assemblage.
This paper highlights current assessment strategies for such older vessels including cubic steel sheet containers as well as cylindrical thick-walled casks. As compliance with each “Konrad”-requirement can hardly be shown independently, only integrated approaches promises success: First, a broader concept of data collection has to be implemented that allows closing documentation gaps by linking dispersed and diverse data, e.g. delivery slips and handwritten fabrication records, by including plau-sibility considerations as well as random inspection programs. Thereby, the relevant properties of all applied casks have to be identified and their potential range should be quantified as accurately as possible. This knowledge is necessary to define one or more representative configurations of the cask as the base for design testing. The actual investigation program must enable to track the influence of deviations and determine safety margins especially in the likely case that not the most conservative setting of all cask features can be tested. The paper evaluates the respective ongoing efforts of industrial applicants and summarizes BAM experiences and associated research projects
Type B(U) packages for the transport of radioactive material have to withstand accident conditions of transport defined in the regulations of the IAEA International Atomic Energy Agency in form of different mechanical (drop) tests with a subsequent thermal test. According to the regulatory requirements the orientation of the package in drop tests shall be such to cause the most damaged state in the components performing the safety functions.
For the package lid system a 9-m drop onto the unyielding target with lid side downwards is often the most damaging orientation. The impact loads acting on the lid in this orientation result mainly from interaction between lid and internal content. In case of a movable content its impact onto the inner side of the lid can cause additional load peaks on the lid and the lid bolts. The intensity of the internal collision depends on the position of content relating to lid at the time of package first contact with target. Due to physical limitations an axial gap, which could be set in “pre-drop” configuration of package or which could spontaneously appear during the drop test, usually does not cover the maximum size possible in specific package design. In this context, the combination of drop tests with post-test analysis can be helpful to better estimate the effect of internal impact.
The paper summarized some aspects of this issue based on the BAM experience in the design assessment of Type B(U) transport packages. Additionally the paper shall support applicants in German approval procedures to reduce rounds of questions and ensure delivery of reliable safety case documents to the authorities. International discussions of this issue at the IAEA and a joint proposal by France and Germany to improve Advisory Material text will be introduced.
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.
Three drop test campaigns have been performed with DN 30 Protective Structural Packaging (PSP) developed by DAHER NUCLEAR TECHNOLOGIES GmbH for the transport of natural, enriched and reprocessed uranium hexafluoride (up to 5 wt%) in 30B cylinders.
The mechanical prototype testing is intended to demonstrate that the package DN30 complies with regulatory requirements under normal and hypothetical accident conditions of transport (NCT, ACT) relevant to IF, AF and B(U)F packages, respectively.
The paper includes the results of the latest test campaign carried out in 2015-2016 at the drop test facility of BAM, Germany with new full scale prototypes of the DN30 PSP and 30B cylinders. Repetition of drop test sequences became necessary after changing to Polyisocyanurate foam as shock absorbing material with variable foam densities. Furthermore, the mechanical behavior of the UF6 content of the 30B cylinder is now simulated by a mixture of cement and steel grid as modified surrogate material; instead of small steel balls. The behavior of this new content simulation is assumed to be more realistic with respect to the properties of real UF6.
BAM as competent authority regarding assessment of mechanical and thermal design, activity release and quality assurance aspects of SNF and HLW transport packages developed a thesaurus of experience feedback topics from recent approval procedures. The list is structured according to the European PDSR guide. It involves issues, which from BAM point of view, needed clarification during last package design assessment procedures. The list contains issues from operation (e.g. deactivation of handling lugs not intended for package handling), maintenance (e.g. leak tightness of impact limiting devices) to technical assessment (e.g. formation of hydrogen by radiolysis and its impact on pressure and ignitability, consideration of ageing mechanisms, thermo-mechanical assessment, impact of gaps between content and flask, spent fuel behaviour) as well as to general and specific safety analysis report requirements (e.g. report structure and required data).
Packages for the transport of radioactive material are often equipped with impact limiters consisting of wood, encapsulated by steel sheets. These impact limiters shall ensure the transport cask meets the mechanical and thermal IAEA regulatory test requirements. After damage caused by the mechanical tests the package has to withstand a severe fire scenario. According to the regulations during and following the thermal test, the specimen shall not be artificially cooled and any combustion of materials of the package shall be permitted to proceed naturally.
Corresponding to results of the French institute IRSN combustion and smouldering of wood inside the impact limiter occurred during and after the fire test. An additional energy supply from a pre-damaged impact limiter to the cask could be the consequence for the safety assessment of the containment.
BAM started a first test phase to examine the issue of combustion for such kind of package components. The goal was to understand the phenomena under the consideration of relevant regulatory boundary conditions. Several metal buckets were filled with wood and equipped with thermocouples. The test specimens have been prepared with different damage arrangements to take the influence of the mechanical tests into account. This paper shows the experimental setup and the conduction of the tests. The first test shows that pre-damaged metal encapsulations can lead to smouldering of the wood and with this to a supplement energy release after the end of the 30 minute fire. BAM is in the preparation process for a second test phase. A thermal test will take place with a wood filled test specimen weighing about 2Mg.
Storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage. Beside consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case. Not only the technical objects are subject of aging but also the safety demonstration basis is subject of “aging” due to possible changes of regulations, standards and scientific/technical knowledge. The basic document, defining the transport safety conditions, is the package design safety report (PDSR) for the transport version of the DPC. To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the PDSR, in connection with periodic renewals of package design approval certificates, have to be carried out. The main reviewing tool is a gap analysis. A gap analysis for a PDSR is the assessment of the state of technical knowledge, standards and regulations regarding safety functions of structures, systems and components.
Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the regulations of the International Atomic Energy Agency. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package capabilities under these conditions. Cracks or failures in the fuel rod cladding can cause the release of gas, volatiles or fuel particles into the cavity. The amount of substances in the cavity has to be considered in the assessment of the activity release and criticality safety.
The mechanical analysis of the compound system formed by the fuel rod cladding and the spent fuel pellets is very difficult due to the limited knowledge of the material properties and the insufficient understanding of the interaction between pellets and cladding and between adjacent pellets. The variation of fuel assembly properties regarding cladding material, burn-up and the history of usage makes reliable predictions of the fuel rod behavior even harder.
For a better understanding about the behavior of spent fuel rods, JRC-ITU and BAM have started a joint research project. In this context, JRC-ITU has developed a test device which allows quasi-static 3-point-bending test on fuel rod segments in the hot cell. The loads are applied with respect to the boundary conditions of the activity release assessment. This paper deals with the numerical calculation of a single fuel rod segment under bending load. The aim is to identify the governing mechanical parameters by the variation of constitutive assumptions, contact conditions, inner constraints, etc. This knowledge helps for the interpretation of the experimental results. Furthermore, the improved understanding about the behavior of the cladding-pellets system will be beneficial for the assessment of spent fuel transport conditions.
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 presents data obtained from experiments performed using a bending test set-up developed at the Joint Research Centre (JRC) – Karlsruhe, for spent fuel segment testing. Adjustable sample holders, loading modes and other experimental conditions can be im- plemented in the experiments to study the effects of different deformation ranges up to cladding failure. The experimental set-up has been adapted to hot cell remote controlling and has a modular configuration, which allows manual and motor-driven loading option. The device has been calibrated on hydrogenated, unirradiated cladding tube segments filled with alumina pellets. The final application of present set-up is to test non-defueled spent fuel rod segments, pressurized to the original spent fuel rod pressure level. The range of applicability of this device, the scope of the experimental program and the first results from actual bending tests will be discussed.
Since 2005, several European countries, coordinated by the European Association of Competent Authorities (EACA), have been developing a guide on contents and structure of the documentation demonstrating the compliance with the regulations for packages for the transport of radioactive material (package design safety report, PDSR). This guide has been periodically improved, considering feedback from Designers and authorities.
Taking into account the successful application of this guide in Europe, in 2013 the International Atomic Energy Agency (IAEA) decided to establish a similar guide as an IAEA document for promotion of worldwide use. The development of this IAEA guide started from the latest version of the European PDSR guide. In 2016/2017, during a 120-day review period, comments on the draft were received from member states and international organizations. These were incorporated into the draft in a series of meetings in 2017.
In another meeting in December 2018 the draft was updated to be in line with the latest revision of the IAEA Regulations for the Safe Transport of Radioactive Material (SSR-6).
In this process the draft has been improved significantly, regarding structure as well as implementation of a graded approach depending on the package type, and clarified.
This paper points to the major considerations in developing the guide and important improvements over the last version of the European PDSR guide.
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.
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 transport package design in accordance with the international IAEA transport regulations as well. The storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage.
Besides consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case.
Not only the technical objects are subject of aging but also the safety demonstration basis is subject of “aging” due to possible changes of regulations, standards and scientific/technical knowledge.
To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the Package Design Safety Report (PDSR), in connection with periodic renewals of package design approval certificates, have to be carried out.
German spent nuclear fuel and high level waste is stored in dry conditions in interim storage facilities until a final repository will be ready for operation in the future. Dual purpose casks are used for the packaging of the materials to ensure transportability at any time during long-term interim storage and fulfilling both requirements from interim storage as well as transport. Germany’s 2011 decision to phase out utilization of nuclear energy has led to a high demand for casks for decommissioning power plants as quickly as possible. Cask serial production will continue at a high level for the next few years. In the process, a system of quality assurance measures ensures that every cask reliably complies with design specification according the transport approval certificate and storage license.
When the first casks were manufactured in Germany about 40 years ago, quality assurance was oriented towards pressure vessel and general nuclear manufacturing. The system was continuously adapted to the state of the art. In particular serial production, globalization of manufacturing and deep supplier chains had to be taken into account. The system of quality assurance measures applied today for cask manufacturing shall be presented here.
The quality assurance measures can be categorized as system-related and design-related. With respect to transport regulations for example, they are independently approved and monitored by the competent authority BAM and its authorized inspection experts. The qualification of the applicant is reviewed first in the context of the design approval procedure and then periodically. The qualification of component and cask manufacturers is also checked at the start of manufacturing and then periodically. Before the start of manufacturing documents such as quality plans are pre-assessed. Manufacturing and inspection are then performed according to the pre-assessed documents. Cask manufacturing is completed by a final inspection before commissioning. Periodic inspections during operation ensure preservation of the properties specified in the approval certificate over package life time. The type and the time interval of periodic inspections depend on the use of the package. For transport after interim storage a system of specific tests and inspections is prescribed.
For disposal of the German research reactor of the Technical University Munich FRM II a new transport and storage cask design was under approval assessment by the German authorities. The Bundesanstalt für Materialforschung und -prüfung (BAM) assessed the mechanical and thermal package safety and performed drop tests. The activity release approaches and subjects of quality assurance and surveillance for manufacturing and operation of the package were assessed by BAM as well.
The cask body is made of ductile cast iron and closed by two bolted lid systems with metal seals. The material of the lids is stainless steel. On each end of the cask a wood-filled impact limiter is installed to reduce impact loads to the cask under drop test conditions. In the cavity of the cask a basket for five spent fuel assemblies is arranged.
For the safety case a combination of experimental testing and analytical/numerical calculations were applied. In total, four drop tests were carried out at the BAM large drop test facility. Two tests were carried out as a full IAEA drop test sequence consisting of a 9m drop test onto an unyielding target and a 1m puncture bar drop test. The other two drop test were performed as single 9m drop tests and completed by additional analyses for considering the effects of an IAEA drop test sequence.
The main objectives of the drop tests were the investigation of the integrity of the package and its safety against release of radioactive material as well as the test of the fastening system of the impact limiters. Furthermore, the acceleration and strain signals measured during the tests were used for the verification of Finite-Element-Analysis (FEA) used for the safety analysis of the package design.
The finite-element models incorporated in the package design safety report include the cask body, the lid system, the inventory and the impact limiters with the fastening system. In this context special attention was paid to the modeling of the encapsulated wood-filled impact limiters.
Additional calculations using the verified numerical models were done by the applicant and assessed by BAM to investigate e.g. the brittle fracture of the cask body made of ductile cask iron within the package design approval procedure.
This paper describes the package design assessment from the view of the competent authority BAM including the applied assessment strategy, the conducted drop tests and the additional calculations by using numerical and analytical methods.
The design of a special form radioactive material has to resist a severe transport accident without undue loss or dispersal of radioactive material.
Safety assessment by authorities competent for design approval has to include besides the required test program (impact, percussion, bending and heat test) also the evaluation of the quality management system for design, manufacture, testing, documentation, use, maintenance and inspection.
These quality assurance measures have to assure that every specimen of the approved design is produced in the same verified quality and every specimen must be able to survive the severe mechanical and thermal tests without undue loss or dispersal of radioactive material at any time of its working life.
All important aspects in the design approval procedure by BAM as the competent authority for approvals of special form radioactive material in Germany are summarized in a guideline published in 2014. This paper will give additional explanations to some aspects in safety assessment, e.g.:
the applicability of leak test methods and the need to consider ageing aspects.
Es wird über die Weiterentwicklung des ASME (American Society of Mechanical Engineers) – Regelwerks berichtet. Relevant für das Design von Behältern ist ASME BPVC (ASME Boiler and Pressure Vessel Code) Section III Division 3. In den ASME BPVC in der Fassung von 2015 wurden dehnungsbasierte Bewertungskriterien aufgenommen. Eine dehnungsbasierte Bewertung erfordert die genaue Kenntnis der aufgetretenen Dehnungen, d.h. speziell möglichst genaue numerische Berechnungen der Behälterbeanspruchungsszenarien. Daher sind regulatorische Vorgaben notwendig für die Vorbereitung, Durchführung und Auswertung numerischer Simulationen. In einer Arbeitsgruppe (ASME Special Working Group on Computational Modeling for Explicit Dynamics) werden dazu Leitlinien für die Anwendung numerischer Berechnungsverfahren entwickelt. Im Vortrag werden diese Leitlinien vorgestellt.
German packages for the transport of spent nuclear fuel are assessed with respect to specific transport conditions which are defined in the safety regulations of the International Atomic Energy Agency.
In general, gastight fuel rods constitute the first barrier of the containment system. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package safety under transport conditions. The objective of this paper is to discuss the methodologies accepted by German authority BAM for the evaluation of spent fuel behavior within the package design approval procedure.
Specific test conditions will be analyzed with regard to assumptions to be used in the activity release and criticality safety analysis. In particular the different failure modes of the fuel rods, which can cause release of gas, volatiles, fuel particles or fragments, have to be properly considered in these assumptions.
The package as a mechanical system is characterized by a complex set of interactions, e.g. between the fuel rods within the assembly as well as between the fuel assemblies, the basket, and the cask containment. This complexity together with the limited knowledge about the material properties and the variation of the fuel assemblies regarding cladding material, burn-up and the operation history makes an exact mechanical analysis of the fuel rods nearly impossible.
The application of sophisticated numerical models requires extensive experimental data for model verification, which are in general not available. The gaps in information concerning the material properties of cladding and pellets, especially for the high burn-up fuel, make the analysis more complicated additionally, and require a conservative approach.
In this context some practical approaches based on experiences by BAM within safety assessment of packages for transport of spent fuel will be discussed.
Ongoing research activities to investigate SNF mechanical behavior in view of gas and fissile material release under transport loads are presented.
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.
Packages for the transport of radioactive material are often equipped with impact limiters consisting of wood, encapsulated by steel sheets. These impact limiters shall ensure the transport cask meets the mechanical and thermal IAEA regulatory test requirements. After damage caused by the mechanical tests the package has to withstand a severe fire scenario. According to the regulations during and following the thermal test, the specimen shall not be artificially cooled and any combustion of materials of the package shall be permitted to proceed naturally.
Corresponding to results of the French institute IRSN combustion and smouldering of wood inside the impact limiter occurred during and after the fire test. An additional energy supply from a pre-damaged impact limiter to the cask could be the consequence for the safety assessment of the containment.
BAM started a first test phase to examine the issue of combustion for such kind of package components. The goal was to understand the phenomena under the consideration of relevant regulatory boundary conditions. Several metal buckets were filled with wood and equipped with thermocouples. The test specimens have been prepared with different damage arrangements to take the influence of the mechanical tests into account. This paper shows the experimental setup and the conduction of the tests. The first test shows that pre-damaged metal encapsulations can lead to smouldering of the wood and with this to a supplement energy release after the end of the 30 minute fire. BAM is in the preparation process for a second test phase. A thermal test will take place with a wood filled test specimen weighing about 2Mg.
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.
The safety demonstrations realized by applicants in the case of approval request for the package designs containing enriched UF6 have to take into account some specific technical issues. Concerning the shipment of enriched UF6, the package designs consist in general of a filled 30B cylinder surrounded by an overpack.
The description of the content, considering the UF6 origin, i.e. natural or reprocessed, shall be clearly justified especially when the UF6 isotopic composition exceeds the limits specified in ASTM standards.
Concerning the containment of the UF6, the applicant shall demonstrate in all conditions of transport the leak-tightness of the valve and plug of the cylinders filled with enriched UF6. In this regard, when mechanical justifications are based on numerical calculations, the absence of contact between these components of the cylinder and the internal surfaces of the overpack after the regulatory drop tests shall be shown to respect the IAEA regulations.
Furthermore, the representativeness of the ballast used to simulate the behaviour of the UF6 loaded within the cylinder shall be justified if drop tests are performed. The representativeness of the ballast should also be justified for numerical calculations.
In addition, the applicant shall demonstrate that the melting temperature of the valve and the plug, including the tinned joint, will not be exceeded during the regulatory fire test.
Finally, specific provisions relative to the use of plugs and the maintenance of cylinders should be included in the safety analysis report.
A Memorandum of Understanding (MOU) between Argonne National Laboratory (Argonne) and the German Bundesanstalt für Materialforschung und -prüfung (BAM) was signed in October 2014. Its objectives are to promote cooperation among scientists and specialists at Argonne and BAM and establish a framework for collaboration in advanced technologies associated with the back end of the nuclear fuel cycle. Collaborative activities involving Argonne and BAM may be implemented through the promotion of joint research activities and scientific workshops and conferences; exchange of technical information; and visits by scientists, specialists, and graduate, postgraduate, and Ph.D. students. The MOU was renewed in January 2020 for another five years. Highlights of the progress and achievements are provided in identified topical areas for collaboration on ageing management guidance, storage, transportation and disposal R&D, joint conference activities, and conclude with the benefits of the MOU in fostering Argonne and BAM collaboration.
For stowage and retention during transport, the International Atomic EnergyAgency (IAEA) Transport regulations (SSR-6) only require that the package shall be securely stowed. Additional information is given in the IAEA guidance material (SSG-26), that includes specific load factors that should be considered in the safety demonstrations.
Nevertheless, applicants commonly use other load factors to justify the design of the packaging attachment points. In particular, acceleration values vary between different countries and this may lead to difficulties during the validation of foreign approval certificates.
Discussions with applicants identified that the load factors presented in the IAEA guidance material are not necessarily relevant for the different transport modes. For some modes, the load factors are representative of situations occurring in routine conditions of transport. For other modes, the load factors are more representative of situations occurring in normal or accident conditions of transport.
Furthermore, the origins of the specified values are not clear. In this context, some IAEA TRANSSC members decided to review the stowage guidance. An international working group was constituted in 2013, including representatives of competent authorities, technical support organizations, and transport stakeholders. Several topics were discussed
and many questions were raised during the two year review. For example, some discussions focused on Paper No.1031 2 the conditions which have to be considered for stowage design, both as relevant to the load factors used for strength and fatigue analysis, as well as the criteria which have to be considered for the package attachment points. In addition, related questions on operational aspects were also discussed. Overall, the working group concluded that new guidance material was warranted as input into SSG-26.
As a result of those discussions, the international working group produced new guidance material for stowage in transport, addressing each of these topics. The proposal to modify the IAEA guidance material (SSG-26, Appendix IV), presented by France to the TRANSSC in the 2015 initiated Review cycle of the Regulations, was unanimously accepted and will be implemented in the next edition of the guidance material.
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 interaction between the package lid system and internal Content during mechanical drop testing is a decisive matter in evaluating Impact loads and the safety of the package. In the case of movable contents ist impact onto the inner side of the package lid can cause additional load peaks on the lid and the lid bolts. Some aspects of this issue were discussed on the basis of experimental results from instrumented drop tests with transport casks and on the basis of analytical approaches.
For transport package design and operation according to the IAEA regulations, the package shall be securely stowed and its retention system shall be capable to withstand load conditions of routine transport. The supporting IAEA Advisory Material SSG-26 provides information how to do that.
Up to now package designers in different countries use other load factors for the design of attachment points than those specified in the IAEA guidance material. In particular the acceleration values vary between different countries and lead to difficulties during the validation of foreign approval certificates. Therefore the IAEA started a discussion process to review the existing guidance text. An international working group was constituted in 2013. Representatives came from different stakeholders, e.g. transport operators, competent authorities and modal organizations. The discussions concluded especially on the transport conditions which has to be considered for stowage design, including on the one hand the relevance of the load factors used for strength and fatigue analysis and on the other hand the criteria which have to be considered for the attachment points.
The proposed acceleration values will be compared to those measured during recent multi-modal testing by Sandia National Laboratories that measured the acceleration levels experienced by a spent fuel flask during heavy-haul truck, sea, and rail transport.
The ISO standard 10276 is dealing with the load attachment systems of packages as well. This standard considers the trunnion design, manufacturing and operational aspects. The regular standard revision phase started in 2017. An expert group discussed new state-of-the-art technology, different analysis approaches for strength and fatigue analysis and proposed revised text for the ISO standard for international discussion. The finite-element analysis approach incl. appropriate acceptance criteria are described and referenced.
The paper describes relevant tie-down aspects, gives background argumentation relevant to analysis approaches, and tries to support harmonized application of the revised IAEA guidance material and the future revised ISO standard.
Transport and storage containers for low and midlevel radioactive waste are getting more and more of relevance due to the nuclear phase out decision in 2011. For higher activities Type B(U) approved waste containers will be needed for the shut down and dismantling phases of NPPs. It is expected that large quantities are required in the near future. German waste containers are generally approved for transport and interim storage and are also intended for the final disposal in the Konrad repository. BAM is involved in the authority licensing of transport package designs in Germany. In this context, BAM is responsible for the assessment of safe containment, mechanical, thermal and quality management issues. BAM also operates test facilities and performs drop as well as thermal tests during package licensing procedures.
This paper summarizes our experience in the transport package design assessment of Type B(U) waste containers. A general overview of the approval process, the requirements and approaches BAM applies are described. Some examples are used to illustrate different aspects and technical issues we are addressing during the package assessment. In the first part of the paper the specific design aspects of German waste containers are described. Here, a general overview is given e.g. about closure systems, impact limiter designs, and the handling concepts. Furthermore, the wide range of radioactive content and their physical behavior including the impact on the packages assessment are described. The second part is focused on questions about the licensing and assessment process of German waste containers from the BAM authority point of view. The general approaches for the strategy of demonstration are outlined on the basis of the test conditions according to IAEA Regulations SSR-6. Furthermore, particular issues of the mechanical and thermal assessment with respect to the specific test conditions are discussed. For accident conditions of transport, aspects to be mentioned are e.g. the assessment of the lid bolts, the axial gap applied between content and lid, and the thermal behavior of the wood filled impact limiter after the fire test. However, issues of the assessment for routine and normal conditions of transport are addressed in this paper, too. The intention of this paper is introduce recent approval procedure experience in Germany, describing technical evaluation issues and so reduce rounds of questions during applications.
Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the regulations of the International Atomic Energy Agency. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package capabilities under these conditions. Cracks or failures in the fuel rod cladding can cause the release of gas, volatiles or fuel particles into the cavity. The amount of substances in the cavity has to be considered in the assessment of the activity release and criticality safety.
The mechanical analysis of the compound system formed by the fuel rod cladding and the spent fuel pellets is very difficult due to the limited knowledge of the material properties and the insufficient understanding of the interaction between pellets and cladding and between adjacent pellets. The variation of fuel assembly properties regarding cladding material, burn-up and the history of usage makes reliable predictions of the fuel rod behavior even harder.
For a better understanding about the behavior of spent fuel rods, JRC-ITU and BAM have started a joint research project. In this context, JRC-ITU has developed a test device which allows quasi-static 3-point-bending test on fuel rod segments in the hot cell. The loads are applied with respect to the boundary conditions of the activity release assessment. This paper deals with the numerical calculation of a single fuel rod segment under bending load. The aim is to identify the governing mechanical parameters by the variation of constitutive assumptions, contact conditions, inner constraints, etc. This knowledge helps for the interpretation of the experimental results. Furthermore, the improved understanding about the behavior of the cladding-pellets system will be beneficial for the assessment of spent fuel transport conditions.
BAM is the German Federal Institute for Materials Research and Testing and the competent authority for mechanical and thermal safety assessment of transport packages for spent fuel and high level waste. In context with safety assessment of RAM packages BAM performed numerous drop tests in the last decades. The tests were mostly accompanied by extensive and various measurement techniques especially by instrumented measurements with strain gages and accelerometers.
The procedure of drop testing and the resulting measurement analysis are the main methods to evaluate the safety against mechanical test conditions. Measurement techniques are dedicated to answer questions in regard to the structural integrity of a RAM package, the mechanical behavior of the prototype as well as of its content under impact conditions.
Test results like deceleration-time functions constitute a main basis for the validation of assumptions in the safety analysis and for the evaluation of numerical calculations. In this context the adequate selection of accelerometers and measurement systems for the performance of drop tests is important. Therefore it is not only necessary to find suitable positions for the accelerometers at the test specimens, but also to consider technical boundary conditions as e.g. temperature.
The Bundesanstalt für Materialforschung und –prüfung (BAM) is a federal institute for materials research and testing in Germany and has been involved in the qualification and safety evaluation procedures of metal seals from the early beginning of the interim storage licensing procedures for radioactive materials, stored in dual purpose casks. Regarding this subject, BAM investigates the long-term behavior of metal seals under the influence of temperature using experimental data and analytical approaches. The development of numerical models is in progress as well. Systematic experimental investigations performed by BAM indicate a continuous decrease of the remaining seal force and the usable resilience considering the leak tightness. Hence, there is a fundamental interest of describing time and temperature dependency to gain predictable values for the long-term behavior and to achieve reliable results with help of short-term tests.
The paper gives an overview about the sealing principle, test program and test results of metal seals of the type HELICOFLEX® HN200. The aging effect, respectively the long-term behavior in dependency of time and temperature, are introduced for two different outer liner materials, aluminum and silver.
Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the regulations of the International Atomic Energy Agency. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important Inputs for the evaluation of the package capabilities under These conditions. Cracks or failures in the fuel rod cladding can cause the release of gas, volatiles or fuel particles into the cavity. The amount of substances in the cavity has to be considered in the assessment of the activity release and criticality safety.
The mechanical analysis of the compound system formed by the fuel rod cladding and the spent fuel pellets is very difficult due to the limited knowledge of the material properties and the insufficient understanding of the interaction between pellets and cladding and between adjacent pellets. The variation of fuel assembly properties regarding cladding material, burn-up and the history of usage makes reliable predictions of the fuel rod behavior even harder.
For a better understanding about the behavior of spent fuel rods, JRC and BAM have started a joint research project. In this context, JRC has developed a test device which allows quasi-static 3-point-bending test on fuel rod segments in the hot cell. The loads are applied with respect to the boundary conditions of the activity release assessment. This paper deals with the numerical calculation of a single fuel rod segment under bending load. The aim is to identify the governing mechanical parameters by the variation of constitutive assumptions, contact conditions, inner constraints, etc. This knowledge helps for the interpretation of the experimental results. Furthermore, the improved understanding about the behavior of the cladding-pellets system will be beneficial for the assessment of spent fuel transport conditions.
Damaged spent nuclear fuel (DSNF) can be loaded in German dual-purpose casks (DPC) for transport and interim storage. Encapsulations are needed to guarantee a safe handling and a tight closure, separated from the package enclosure. These encapsulations shall be durable and leak-tight for a long storage period, because they are usually not accessible within periodical inspections of the DPC. Due to the general design of DPCs for standard fuel assemblies, specific requirements have to be considered for the design of encapsulations for DSNF to ensure the loading in existing package designs. Especially the primary lid system of a DPC is designed for maximum loads due to the internal impact of the content during drop test conditions. The main difference of encapsulations for damaged spent nuclear fuel is that they have usually a much higher stiffness than standard fuel assemblies. Therefore the design of an internal shock absorber, e.g. at the head of an encapsulation is required to reduce mechanical loads to the primary lid system during impacts.
BAM as part of the German competent authority system is responsible for the safety assessment of the mechanical and thermal package design, the release of radioactive material and the quality assurance of package manufacturing and operation. Concerning the mechanical design of the encapsulation BAM was involved in the comprehensive assessment procedure during the package design approval process. An internal shock absorber was developed by the package designer with numerical analyses and experimental drop tests. Experimental drop tests are needed to cover limiting parameters regarding, e.g. temperature and wall thickness of the shock absorbing element to enable a detailed specification of the whole load-deformation behavior of the encapsulation shock absorber.
The paper gives an overview of the assessment work by BAM and points out the main findings which are relevant for an acceptable design of internal shock absorbers. The physical drop tests were planned on the basis of pre-investigations of the applicant concerning shape, dimension and material properties. In advance of the final drop tests the possible internal impact behavior had to be analyzed and the setup of the test facility had to be validated. The planning, performance and evaluation of the final drop tests were witnessed and assessed by BAM. In conclusion it could be approved that the German encapsulation system for damaged spent nuclear fuel with shock absorbing components can be handled similar to standard fuel assemblies in existing package designs.
BAM is the German Federal Institute for Materials Research and Testing and the competent authority for mechanical and thermal safety assessment of transport packages for spent fuel and high level waste. In context with safety assessment of RAM packages BAM performed numerous drop tests in the last decades. The tests were mostly accompanied by extensive and various measurement techniques especially by instrumented measurements with strain gages and accelerometers. The procedure of drop testing and the resulting measurement analysis are the main methods to evaluate the safety against mechanical test conditions. Measurement techniques are dedicated to answer questions in regard to the structural integrity of a RAM package, the mechanical behavior of the prototype as well as of its content under impact conditions.
Test results like deceleration-time functions constitute a main basis for the validation of assumptions in the safety analysis and for the evaluation of numerical calculations. In this context the adequate selection of accelerometers and measurement systems for the performance of drop tests is important. Therefore it is not only necessary to find suitable positions for the accelerometers at the test specimens, but also to consider technical boundary conditions as e.g. temperature.
Packages for the transport of radioactive material shall meet the mechanical and thermal test requirements of the International Atomic Energy Agency (IAEA) regulations for package design approval. Besides mechanical testing, the Federal Institute for Materials Research and Testing (BAM) performs thermal tests in accordance with the IAEA regulations. The thermal test includes a 30-minute 800°C fully engulfing fire.
BAM continuously performs various thermal experiments for the investigation of the thermal response of packages with respect to the IAEA fire. The purpose of this paper is to give an overview of the already performed, ongoing and future physical tests and experiments of BAM in the field of thermal investigations. These research and development works shall support our competencies for the authority package design assessment.
BAM operates a propane gas fire test facility. To be able to carry out comparative investigations and validity between the propane fire and the in detail prescribed pool fire test in the regulations, BAM carries out various calorimetric tests and investigates the boundary conditions of the fire with the help of fire reference packages. At the same time, we are conducting various fire scenarios with wood-filled impact limiters. Large-scale fire tests of impact limiters are carried out on a full scale as well as on a small scale. Influencing variables are investigated in particular by means of geometric changes and the consideration of artificial damages, in particular holes. In addition to propane fire as a heat source, thermal scenarios are also investigated with hydrogen as heat source and an infrared radiator system to ignite test specimens. For these numerous test arrangements, the transferability to existing and newly developed transport package designs is essential and fruitful within the review of design approvals, especially for Dual Purpose casks with a long-lasting operation time.
Type B(U) packages for the transport of radioactive material have to withstand accident conditions of transport defined in the regulations of the IAEA International Atomic Energy Agency in form of different mechanical (drop) tests with a subsequent thermal test. According to the regulatory requirements the orientation of the package in drop tests shall be such to cause the most damaged state in the components performing the safety functions.
For the package lid system a 9-m drop onto the unyielding target with lid side downwards is often the most damaging orientation. The impact loads acting on the lid in this orientation result mainly from interaction between lid and internal content. In case of a movable content its impact onto the inner side of the lid can cause additional load peaks on the lid and the lid bolts. The intensity of the internal collision depends on the position of content relating to lid at the time of package first contact with target. Due to physical limitations an axial gap, which could be set in “pre-drop” configuration of package or which could spontaneously appear during the drop test, usually does not cover the maximum size possible in specific package design. In this context, the combination of drop tests with post-test analysis can be helpful to better estimate the effect of internal impact.
The paper summarized some aspects of this issue based on the BAM experience in the design assessment of Type B(U) transport packages. Additionally the paper shall support applicants in German approval procedures to reduce rounds of questions and ensure delivery of reliable safety case documents to the authorities. International discussions of this issue at the IAEA and a joint proposal by France and Germany to improve Advisory Material text will be introduced.
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.
Based on state-of-the-art research and regulations, the application of the fracture mechanics master curve (MC) concept to ferritic ductile cast iron (DCI) is being investigated in a recently started German joint research project between BAM Berlin and MPA Stuttgart. This paper outlines the research approach and the planned investigations and provides first results.
According to the IAEA guidelines SSG 26, the safety assessment of DCI containers for transport as well as interim and final storage of radioactive materials is based on the fracture mechanics criterion of general crack initiation exclusion. Within this context, the focus of the current research project is on the further development of brittle fracture safety assessment methods under dynamic loading conditions. Within fracture mechanics safety assessments for steels in nuclear technology, the probabilistic MC concept according to ASTM E1921 is currently used for quasi-static loading conditions as a supplement to the established deterministic ASME reference curve concept. The MC concept allows for an effective statistical consideration of the scatter of the material toughness in the transition regime. Furthermore, the MC can also be used for elevated loading rates as they occur in accident or crash scenarios. However, for DCI, a systematic review of potential modifications to the MC assumptions and procedure according to ASTM E1921 and an associated validation are still lacking. Therefore, a corresponding methodology shall be established to determine and assess dynamic fracture toughness values of DCI in the ductile-to-brittle transition regime, using samples extracted from a component.
For the experimental program, a DCI material was produced as it is generally used for transport and storage containers. A basic mechanical-technological material characterization consisting of tensile, Charpy, and Pellini tests is planned. A comprehensive fracture mechanics test program was compiled covering dynamic testing of DC(T)9, C(T)25, C(T)50, SE(B)10, SE(B)25, and SE(B)140 specimens. Moreover, numerical analyses of different specimen geometries and test setups are planned to investigate the influences of stress triaxiality and loading rate. Throughout the experimental program, the microstructure-property relationship will be investigated based on quantitative metallographic and fractographic analyses. Finally, different MC assessment strategies will be considered, and modifications will be made if necessary.
Piezoresistive accelerometers use a strain-sensing element, generally made of semiconductor material, e.g., silicon to convert the mechanical motion into an electrical signal. This element is usually designed in form of a cantilever beam loaded with a mass. Acceleration causes bending of the beam, which produces a change of electrical resistance proportional to the applied acceleration.
Main advantages of piezoresistive accelerometers in comparison to other types, e.g., piezoelectric and capacitive, is their robust and highly dynamic behavior, which qualifies them for application in high impact shock applications. Mechanical damping is typically implemented with silicon oil in a way that the output signal is undistorted over a wide frequency range. These characteristics principally qualify them for the application in drop tests carried out at BAM, for which they are calibrated over the frequency range from 1 to 4 kHz. However, using silicon oil for damping, has the drawback of temperature dependent change of its viscosity, leading to temperature dependent deviation of the accelerometer’s sensitivity.
This study presents experimental results of the temperature behavior of a piezoresistive accelerometer with a dynamic range up to ±5000 g.
This type of accelerometer is applied for drop tests which are partially performed at temperatures of -40 or +100 °C.
In context with Japanese disposal and transport container safety assessment the German Federal Institute for Materials Research and Testing (Bundesanstalt für Materialforschung und -prüfung) per-formed extensive drop test series with prototypes of RAM packages in recent years.
The paper presents two cooperative projects focused on project management, testing performance and experimental results.
In the first project, a full-scale prototype of a package for transport and storage of spent fuel elements with a total mass of 127,000 kg and its reduced-scale model which was similar to the prototype in both geometry and design were drop tested in 9 m and 1 m puncture tests at various drop orientations under regulatory conditions. In context with a corresponding research project the test data was used in combination with numerical methods to investigate the phenomenon of internal cask- content collisions during drop tests. Here, technological gaps between basket or radioactive content and cask body or primary lid can be a reason for significant high dynamic loadings of cask components and its internals due to additional impact interactions in the cask cavity caused by inner relative movement between content and cask. Further investigations were carried out in the field of similarity mechanics comparing experimental drop test data of the full-scale prototype and reduced-scale model in regard to the transferability of package impact response from reduced-scale models to full-scale packages.
In the second project, drop tests with a waste container for intermediate depth disposal were per-formed being contracted by Kobe Steel, Ltd. and a consortium of Japanese electric power plant com-panies. The drop test program comprised three single 8-m drop tests in a specimen’s corner edge orientation. For drop tests according to legal regulations for interim storage or final disposal it is often necessary to use a target which represents the real ground of the storage facility. In such cases the container hits directly onto a well-defined concrete slab on top of the IAEA target. Here, the directly impact target for the drop tests were concrete slabs manufactured in Japan and connected by mortar to the impact pad of the unyielding IAEA target.
Temperature characteristics of a piezoresistive accelerometer for high impact shock application
(2018)
This study presents the characterization of a piezoresistive accelerometer damped with silicon oil for the application in drop tests carried out at BAM. Experiments were performed with the Hopkinson Bar method in close correlation to the real-world application conditions. The results point out certain limitations regarding the temperature influence and the frequency response. Additional experiments were performed with a gas damped type of piezoresistive accelerometer, which has superior specifications, particularly for low temperatures. The results allow for a comparison.
The first German package design approval certificate for a dual purpose cask intended for loading with damaged spent nuclear fuel was issued recently. BAM as part of the competent authority system in Germany carried out a comprehensive assessment procedure with respect to the mechanical and thermal design, the release of radioactive material and the quality assurance aspects of manufacturing and operation. Packages for the transport and storage of radioactive material have been assessed by BAM for many years, thus the common assessment procedure is well-known and good practice. Up to now only SNF without defects or HLW with well-defined properties were designated for long-term Interim storage and transports afterwards. Due to Germany’s nuclear phase out all other kinds of spent nuclear fuel in particular damaged spent nuclear fuel shall be packed as well. Damaged spent nuclear fuel needs a tight closure with Special encapsulations and clearly defined properties in Germany. In addition, these encapsulations shall be long-term durable, because they are not accessible after loading in a packaging within periodical inspections. The main difference to Standard package components is that encapsulations with a permanent closure achieve their specified conditions not after manufacturing but only during operation, after loading and closing. To ensure compliance with the specific conditions, special measures for quality assurance are necessary during operation of each encapsulation, e.g. drying and sealing, which were assessed by BAM. The present paper gives an overview of the conducted assessment from BAM and point out the findings concerning to the special closure lid of the approved encapsulation, which is screwed and welded. A wide verification concept is necessary to show the specific tightness under transport conditions. Together with quality assurance measures during first operation steps these encapsulations with damaged spent nuclear fuel can be handled like standard fuel assemblies in approved package designs.
German package design approvals were granted recently for dual purpose casks (DPC) intended for loading with encapsulated damaged spent nuclear fuel (DSNF). Comprehensive assessment procedures were carried out by the authority BAM with respect to the mechanical and thermal package design, the activity release of radioactive material and quality assurance aspects for manufacturing and operation of each packaging. The objective of each procedure was to verify the Package Design Safety Report (PDSR) and the relevant guidelines fulfils the requirements of the IAEA regulations.
Previous approvals of German SNF package designs consider mainly standard fuel assemblies with defined specifications and properties for transport and interim storage. Due to the nuclear power phase-out in Germany all kinds of SNF, e.g. damaged spent fuel rods shall be packed in DPC now. Therefore specific requirements shall be considered in accordance with international experiences including IAEA technical reports. The main requirement for DSNF is a tight encapsulation with specific defined properties under transport and storage conditions.
Due to the interim storage period of currently up to 40 years the encapsulation with DSNF in the casks shall also be long term durable. Thus specific loading and drying procedures are necessary and had to be qualified during the approval process. BAM assessed these drying procedures and could confirm the long-term behaviour of the encapsulation and the suitability of the drying equipment. This special equipment was qualified in a “cold handling”. In addition, it was shown that the behaviour of the test equipment used in the qualification process was comparable with the original equipment, e.g. test fuel rods or test encapsulation. In the development of the drying process, experience was obtained in how to put the requirements of the IAEA regulations and related IAEA technical reports into practice.
The paper gives an overview of approval assessment and testing experience made by BAM and point out the main resulting requirements on drying processes for these kinds of encapsulations with DSNF.
For disposal of the research reactor of the Technical University Munich FRM II a new transport and storage cask design was under approval assessment by the German authorities on the basis of International Atomic Energy Agency (IAEA) requirements. The cask body is made of ductile cast iron and closed by two bolted lid systems with metal seals. The material of the lids is stainless steel. On each end of the cask the wood-filled impact limiters are installed to reduce impact loads to the cask under drop test conditions. In the cavity of the cask a basket for five spent fuel elements is arranged.
This design has been assessed by the Bundesanstalt für Materialforschung und -prüfung (BAM) in view to the mechanical and thermal safety analyses, the activity release approaches, and subjects of quality assurance and surveillance for manufacturing and operation of the package.
For the mechanical safety analyses of the package a combination of experimental testing and analytical/numerical calculations were applied. In total, four drop tests were carried out at the BAM large drop test facility. Two tests were carried out as a full IAEA drop test sequence consisting of a 9m drop test onto an unyielding target and a 1m puncture bar drop test. The other two drop tests were performed as single 9m drop tests and completed by additional analyses for considering the effects of an IAEA drop test sequence.
The main objectives of the drop tests were the investigation of the integrity of the package and its safety against release of radioactive material as well as the test of the fastening system of the impact limiters. Furthermore, the acceleration and strain signals measured during the tests were used for the verification of finite-element (FE) models applied in the safety analysis of the package design.