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Abstract Hydrogenated nitrile butadiene rubber (HNBR) elastomer was thermo-oxidatively aged at different temperatures up to 150 °C. Fourier transform infrared spectroscopy (FTIR), compression stress relaxation (CSR) and international rubber hardness degree (IRHD) microhardness were used to characterise the chemo-mechanical changes of HNBR O-rings during thermo-oxidative ageing. FTIR shows the development of carbonyl, methyl and ester groups but the nitrile content was not affected by ageing. The effect of sample geometry during CSR was investigated. CSR data were converted through integrated kinetic laws. The conversion has proven its sensibility to detect heterogeneous ageing. This was confirmed by the IRHD measure-ments across the section of O-rings. The influence of compression during ageing was assessed through IRHD measurements across the section of compressed and uncom-pressed aged O-rings. The DLO effect was more pronounced in compressed O-rings. By applying the model of Wise et al., theoretical IRHD and oxidation profiles were determined on the basis of IRHD experimental data of compressed O-rings. Good agreements between the experimental and the theoretical IRHD profiles in the core region were obtained. However, near the edge, the theoretical IRHD values were overestimated.
EPDM, HNBR and FKM materials were exposed at 150 ◦C to air under atmospheric pressure and to hydrogen at 50 bar for different ageing times. All measurements after hydrogen exposure were conducted on samples in degassed condition to assess irreversible effects resulting from that exposure and to compare them to those after ageing in air. Density, hardness, tensile properties, compression set, and hydrogen permeability of all samples were analysed. In both ageing environments, HNBR exhibited the most significant changes of material properties.
However, for both EPDM and HNBR, considerably less severe ageing effects were observed under hydrogen in comparison to ageing in air. On the other hand, FKM showed about the same low level of deterioration in both ageing environments but exhibited poor resistance against damage due to rapid gas decompression in hydrogen environment that can lead to seal failure. The obtained results may serve as a guidance toward a better understanding for design and utilisation of elastomeric materials in future hydrogen infrastructure components.
A state-of-the-art (SOTA) review on characterisation of spent nuclear fuel (SNF) properties in terms of source term and inventory assessment (neutron, gamma-ray emission, decay heat, radionuclide inventory, elemental content) and in terms of out-of-core fuel performance (cladding performance and fuel integrity in view of the safety criteria for SNF interim storage, transport and canister packaging) using several numerical and experimental approaches and methodologies is presented. This SOTA report is a result of the spent fuel characterisation (SFC) work package as part of the European Joint Programme on Radioactive Waste Management (EURAD), which offers an overview of the status of knowledge in the field of SNF characterisation and assessment during the pre-disposal phase. The document aims to focus on the current safety-significant gaps and related challenges, providing a direct link to the goals of the mandated actors of EURAD. The report is expected to be used by all EURAD colleagues in their national programmes as a key resource for knowledge management programmes and to contribute to demonstrating and documenting the state-of-the-art.
The prevention of brittle fracture of spent fuel claddings during long-term dry interim storage is based on experimental investigations, numerical analyses and assessment methods for predicting the mechanical behavior and determining limiting conditions. The ring compression test (RCT) is an established experimental method for characterizing cladding material. Test results for various high-burnup pressure water reactor zirconium-based fuel cladding alloys (e.g., ZIRLO®, M5®) are publicly available. To reduce the effort associated with irradiated samples in hot cells, it is helpful to perform studies on unirradiated surrogate cladding material. Based on such experimental data, load-displacement curves were numerically analyzed for selected cladding materials. In the presence of radial hydrides, a sample may suddenly fail by fracture even at small deformations. Noticeable load drops in the RCT occur associated to unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting neighboring hydrides. The failure process was simulated by cohesive zones controlled by the fracture energy and the cohesive strength. A modeling approach is presented in which the radial hydride morphology is taken into account. Based on the developed fracture mechanics approach with cohesive zone modeling, not only the deformation behavior but also the failure behavior of irradiated as well as unirradiated claddings with radial hydrides under RCT loading conditions can be adequately described.
Task 3 of Work package 8 (Spent Fuel Characterization) of the European Joint Programme on Radioactive Waste Management (EURAD) investigates the behaviour of nuclear fuel and cladding after discharge. The aim of the work is to understand and describe the be-haviour of spent nuclear fuel (SNF), irradiated cladding, fuel/cladding chemical interaction (FCCI) and ageing under conditions of extended interim storage, transportation and em-placement in a final disposal system. BAM contributes to the project as partner and leads Task 3. The presentation gives an overview of the project status, main achievements in experimental work and modelling studies, deviations from the plan, delays and challenges ahead.
The prevention of brittle fracture of spent fuel claddings during long-term dry interim storage is based on experimental investigations, numerical analyses and assessment methods for predicting the mechanical behavior and determining limiting conditions. The ring compression test (RCT) is an established experimental method for characterizing cladding material. Test results for various high-burnup pressure water reactor zirconium-based fuel cladding alloys (e.g., ZIRLO®, M5®) are publicly available. To reduce the effort associated with irradiated samples in hot cells, it is helpful to perform studies on unirradiated surrogate cladding material. Based on such experimental data, load-displacement curves were numerically analyzed for selected cladding materials. In the presence of radial hydrides, a sample may suddenly fail by fracture even at small deformations. Noticeable load drops in the RCT occur associated to unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting neighboring hydrides. The failure process was simulated by cohesive zones controlled by the fracture energy and the cohesive strength. A modeling approach is presented in which the radial hydride morphology is taken into account. Based on the developed fracture mechanics approach with cohesive zone modeling, not only the deformation behavior but also the failure behavior of irradiated as well as unirradiated claddings with radial hydrides under RCT loading conditions can be adequately described.
In this work we present results for three typical sealing materials (EPDM, HNBR, FKM) after up to 100 days ageing at high temperature under hydrogen. To allow for a better evaluation of the occurring changes the results are compared with effects of ageing in air on the material properties. Despite the common assumption that ageing under hydrogen atmosphere is expected to be less severe for the material in com-parison to air ageing, this is not seen for every material.
The storage of spent fuel and high level radioactive waste in Germany is performed in interim storage containers with double lid systems. The lids are bolted and equipped with metal seals (e.g. Helicoflex®) to ensure the safe enclosure of the inventory. The used metal seals have a layered structure consisting of three components as can be seen schematically in the cross-sectional view in Fig. 1.
In the center a helical spring is positioned that is surrounded by two C-shaped jackets and is mainly responsible for generation of the required restoring force. The inner jacket is made of stainless steel and homogenizes the restoring force of the helical spring. The outer jacket is made of silver or aluminium which both are soft metals in comparison to the contact partners (lid and container body). During bolting of the lid to the container body the seal is compressed. The generated restoring force of the helical spring causes a plastic deformation of the outer jacket and adapts to the surfaces of the lid and the container body. Hence, leakage paths are closed and the sealing function is generated.
Typical durations for existing interim storage licenses in Germany are 40 years, but it can be expected that they have to be extended to longer periods as a final repository will not be available before the end of the running licence periods. This extension of license periods requires a solid understanding of the long-term behaviour of the seals under storage conditions. To meet this challenge long-term investigations have been started at Bundesanstalt für Materialforschung und –prüfung (BAM) in 2009. These tests focus on seals assembled in test flanges which are stored at temperatures ranging from room temperature to 150 °C for accelerated ageing. The aged seals are tested repeatedly after certain ageing steps and the leakage rate as indicator for sealing performance, the remaining seal force, and the useable resilience upon decompression are determined.
In the poster an update on the performed investigations in respect to earlier publications (Grelle et al. 2019, Goral et al. 2023) will be given and the implications of the results for resilient long term safety will be presented. Additionally, a focus will be laid on the currently planned further investigations and the question “What is additionally needed for evaluation of an interim storage period extension in regard to the used metal seals?” will be addressed.
Beginning in 1976 the former iron ore mine Konrad was geologically investigated concerning its suitability as a location for a German final repository for low and intermediate level waste. After a comprehensive licensing procedure it was approved and is now being prepared and equipped for the planned start of storage in 2027. At the end of the 1980s, the requirements for waste containers for storage in this repository were defined for the first time on the basis of an incident analysis of the on-site handling procedures, beginning with the unloading of the packages after the arrival at the site up to the final positioning in the underground storage galleries. A 5-meter drop onto the rock ground of the mine was identified as the covering case for high mechanical requirements (ABK II container class). In contrast to the 9-meter drop according to the IAEA Regulations for the Safe Transport of Radioactive Material, the 5-meter drop is performed onto a hard but not essentially unyielding target. The container is typically not protected by an impact limiter. The requirements for a potential test facility are described in the regulations for the Konrad repository. The mechanical strength of the target is defined as a concrete strength equivalent to the identified properties of the rock ground of the mine.
Since 1991 BAM has consistently used precast reinforced concrete slabs as target for drop tests in the framework of licensing procedures as well as in research projects. While the original design fulfilled the requirement for the integrity of the concrete slab in most cases, it failed when drop tests onto an edge of a container were performed. A redesigned concrete slab developed in a research project and suggested as a reference target in 2009 has been successfully used in Konrad licensing procedures since then.
The paper gives a brief overview of the historical development and it describes BAM’s efforts and approaches to continuously guarantee a concrete slab of defined quality and to provide a test setup for valid drop tests from the applicants and the authorities view.
Beginning in 1976 the former iron ore mine Konrad was geologically investigated concerning its suitability as a location for a German final repository for low and intermediate level waste. After a comprehensive licensing procedure it was approved and is now being prepared and equipped for the planned start of storage in 2027. At the end of the 1980s, the requirements for waste containers for storage in this repository were defined for the first time on the basis of an incident analysis of the on-site handling procedures, beginning with the unloading of the packages after the arrival at the site up to the final positioning in the underground storage galleries. A 5-meter drop onto the rock ground of the mine was identified as the covering case for high mechanical requirements (ABK II container class). In contrast to the 9-meter drop according to the IAEA Regulations for the Safe Transport of Radioactive Material, the 5-meter drop is performed onto a hard but not essentially unyielding target. The container is typically not protected by an impact limiter. The requirements for a potential test facility are described in the regulations for the Konrad repository. The mechanical strength of the target is defined as a concrete strength equivalent to the identified properties of the rock ground of the mine.
Since 1991 BAM has consistently used precast reinforced concrete slabs as target for drop tests in the framework of licensing procedures as well as in research projects. While the original design fulfilled the requirement for the integrity of the concrete slab in most cases, it failed when drop tests onto an edge of a container were performed. A redesigned concrete slab developed in a research project and suggested as a reference target in 2009 has been successfully used in Konrad licensing procedures since then.
The paper gives a brief overview of the historical development and it describes BAM’s efforts and approaches to continuously guarantee a concrete slab of defined quality and to provide a test setup for valid drop tests from the applicants and the authorities view.
Beginning in 1976 the former iron ore mine Konrad was geologically investigated concerning its suitability as a location for a German final repository for low and intermediate level waste. After a comprehensive licensing procedure it was approved and is now being prepared and equipped for the planned start of storage in 2027. At the end of the 1980s, the requirements for waste containers for storage in this repository were defined for the first time on the basis of an incident analysis of the on-site handling procedures, beginning with the unloading of the packages after the arrival at the site up to the final positioning in the underground storage galleries. A 5-meter drop onto the rock ground of the mine was identified as the covering case for high mechanical requirements (ABK II container class). In contrast to the 9-meter drop according to the IAEA Regulations for the Safe Transport of Radioactive Material, the 5-meter drop is performed onto a hard but not essentially unyielding target. The container is typically not protected by an impact limiter. The requirements for a potential test facility are described in the regulations for the Konrad repository. The mechanical strength of the target is defined as a concrete strength equivalent to the identified properties of the rock ground of the mine.
Since 1991 BAM has consistently used precast reinforced concrete slabs as target for drop tests in the framework of licensing procedures as well as in research projects. While the original design fulfilled the requirement for the integrity of the concrete slab in most cases, it failed when drop tests onto an edge of a container were performed. A redesigned concrete slab developed in a research project and suggested as a reference target in 2009 has been successfully used in Konrad licensing procedures since then.
The paper gives a brief overview of the historical development and it describes BAM’s efforts and approaches to continuously guarantee a concrete slab of defined quality and to provide a test setup for valid drop tests from the applicants and the authorities view.
Cohesive Zone Modelling Approach on Irradiated Claddings Subjected to Long-Term Dry Interim Storage
(2023)
Long-term dry interim storage may adversely affect the mechanical properties of spent fuel rods, possibly resulting in a reduced resilience during handling or transport after storage. Pre-storage drying and the early stage of interim storage can subject the cladding to higher temperatures and higher pressure induced tensile hoop stresses than those associated with in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate in zirconium-based alloys (Zircaloy) during slow cooling, which may result in embrittlement of the cladding material and eventually a sudden failure of cladding under additional mechanical loads. Especially long, continuous radial hydride structures and low temperature can cause severe embrittlement of claddings and finally failure by fracture even at small deformations. The focus of the presented research is on the development of appropriate numerical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. An iterative inverse analysis method is used for deriving the elastic-plastic material properties in the hoop direction of a ring-shaped sample. A modelling approach based on cohesive zones is explained which can reproduce the propagation of cracks initiated at radial hydrides in the zirconium matrix. The developed methods are applied to defueled samples of cladding alloy ZIRLO®, which were subjected to a thermo-mechanical treatment to reorient existing circumferential hydrides to radial hydrides. A selected sample showing sudden load drops during a quasi-static ring compression test is analysed by means of fracture mechanics for illustrative purposes.
The European Joint Programme on Radioactive Waste Management (EURAD) is working on Spent Fuel Characterization (SFC) in its work package (WP) 8. Inspired by the EURAD activities, the International Atomic Energy Agency (IAEA) established an international Coordinated Research Project (CRP) on SFC. The EURAD WP SFC
participants are collaborating as a team on the IAEA CRP on SFC. The EURAD WP SFC project consists of four tasks. Task 3 investigates the behaviour of nuclear fuel and cladding after discharge. The aim of these activities is to understand and describe the evolution of the cladding-pellet system and its ageing under conditions of extended interim storage, transportation and emplacement in a final disposal system. At a Consultancy Meeting, BAM as contributor to Task 3 presented current results of the failure analysis of irradiated ZIRLO® claddings under conditions of the Ring Compression Test.
Elastomeric seals are essential components in the infrastructure which prevent leakage of gas and proper function of technical devices and are therefore highly safety relevant. For proper function a remaining resilience and tolerance to pressure changes is required.
The ageing of elastomers is an undesirable but inevitable process leading to a limited lifetime of rubber products such as seals. Independent of the application, as e.g. automotive, piping or container applications, a long lifetime or at least a timely exchange of a seal before occurrence of critical failure is desired.
Hydrogen is a key element for the energy turnaround and therefore the compatibility of seals with hydrogen is essential. In addition to the effect of hydrogen uptake and release, which might also cause destructive effects such as rapid gas decompression, ageing at high temperature under hydrogen atmosphere is of interest too. Most of the existing work is addressing the performance of new materials and comprises only very limited investigations concerning long-term use and the behaviour of aged materials in contact with hydrogen. As ageing can lead to substantial changes of material properties, it must be evaluated whether these changes are beneficial or deteriorating for the component function.
In this work we present and compare results of the characterisation of three sealing materials (EPDM, HNBR, FKM) after ageing at high temperature under hydrogen and air. Despite the common assumption that ageing under hydrogen atmosphere should be less severe for the material in comparison to air ageing, this is not the case for every material.
Overview of ongoing research and future prospects on polyethylene neutron shielding materials at BAM
(2023)
The extension of the interim storage period of radioactive waste before disposal will cause additional challenges for the nuclear waste management in Germany, so that an extensive knowledge of the long-term performance of casks, including their components and inventories, will be required for future extended storage licenses.
Ultra-high and high molecular weight polyethylenes ((U)HMW-PE) are used for neutron shielding purposes in casks for storage and transport of spent fuel and high-level waste due to their extremely high hydrogen content. During their service life of several decades as cask components, the PE materials are exposed to neutron and gamma radiation from the radioactive inventory of the casks, mechanical assembling stresses and temperature. All these combined effects affect the material properties of such components which in turn may be crucial for some possible accident scenarios.
At the Bundesanstalt für Materialforschung und -prüfung (BAM), the effects of high temperature exposure in combination with subsequent or previous irradiation were investigated with a comprehensive aging program including thermal aging at 125 °C for different aging periods up to 5 years and irradiation with doses ranging from 50 to 600 kGy.
This contribution provides an overview of the ongoing research related to the structural changes of (U)HMW-PE induced by gamma irradiation and high temperature exposure and focuses on current research perspectives at BAM with regard to the prediction of the dynamic behavior of the material during extended interim storage in case of an accident scenario. First results of the coupled effect of temperature, radiation and mechanical loading will be presented. The effect of microstructural changes induced by gamma irradiation and high temperature on the mechanical behavior of (U)HMW-PE will be assessed.
Overview of ongoing research and future prospects on polyethylene neutron shielding materials at bam
(2023)
The extension of the interim storage period of radioactive waste before disposal will cause additional challenges for the nuclear waste management in Germany, so that an extensive knowledge of the long-term performance of casks, including their components and inventories, will be required for future extended storage licenses.
Ultra-high and high molecular weight polyethylenes ((U)HMW-PE) are used for neutron shielding purposes in casks for storage and transport of spent fuel and high-level waste due to their extremely high hydrogen content. During their service life of several decades as cask components, the PE materials are exposed to neutron and gamma radiation from the radioactive inventory of the casks, mechanical assembling stresses and temperature. All these combined effects affect the material properties of such components which in turn may be crucial for some possible accident scenarios.
At the Bundesanstalt für Materialforschung und -prüfung (BAM), the effects of high temperature exposure in combination with subsequent or previous irradiation were investigated with a comprehensive aging program including thermal aging at 125 °C for different aging periods up to 5 years and irradiation with doses ranging from 50 to 600 kGy.
This contribution provides an overview of the ongoing research related to the structural changes of (U)HMW-PE induced by gamma irradiation and high temperature exposure and focuses on current research perspectives at BAM with regard to the prediction of the dynamic behavior of the material during extended interim storage in case of an accident scenario. First results of the coupled effect of temperature, radiation and mechanical loading will be presented. The effect of microstructural changes induced by gamma irradiation and high temperature on the mechanical behavior of (U)HMW-PE will be assessed.
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.
Long-term dry interim storage may adversely affect the mechanical properties of spent fuel rods, possibly resulting in a reduced resilience during handling or transport after storage. Since the cladding is the first barrier for the spent fuel pellets, its integrity must be demonstrated until the end of interim storage and subsequent transportation. An established method for characterizing the cladding material is the ring compression test, in which a small, cylindrical sample of the cladding tube is subjected to a compressive load. This test is a laboratory representation of a load case where the fuel rod is crushed.
Pre-storage drying and the early stage of interim storage can subject the cladding to higher temperatures and higher pressure induced tensile hoop stresses than those associated with in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate in zirconium-based alloys (Zircaloy) during slow cooling, which result in embrittlement of the cladding material and eventually a possible sudden failure of cladding integrity under additional mechanical loads. Especially long, continuous radial hydride structures and low temperature can cause severe embrittlement of claddings and finally failure by fracture even at small deformations. Therefore, the study of hydride morphology plays an important role in describing the brittle failure behaviour of the claddings.
The focus of the presented research is on the development of appropriate numerical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. Typical hydride morphologies are shown. An iterative inverse analysis method is described for deriving the elastic-plastic material properties in the hoop direction of a ring-shaped sample. A modelling approach based on cohesive zones is explained which is able to reproduce the propagation of cracks initiated at radial hydrides in the zirconium matrix. The developed methods are applied to defueled samples of cladding alloy ZIRLO®, which were subjected to a thermo-mechanical treatment to reorient existing circumferential hydrides to radial hydrides. A selected sample showing sudden load drops during a quasi-static ring compression test is analysed by means of fracture mechanics for illustrative purposes.
This project as part of the European Joint Programme on Radioactive Waste Management has received funding from the European Union’s Horizon 2020 research and innovation programme under grant agreement no. 847593.
It is common practice in the application of finite element analysis to model compression set (CS) of elastomers during aging with two different material models according to the two-network theory of Tobolsky. The theory relies on the existence of two networks. The first one represents the original network after vulcanization and is sensitive to chain scission. The second network accounts for the formation of additional crosslinking during aging. Besides the use of user subroutines to describe the two-network model, an element overlay technique is also needed as the full set of both material behaviors did not exist for assignment to a single element. This element overlay technique is valuable for research and developmental purposes but makes extension to industrial usage quite challenging. Our goal is to simulate the CS of elastomers after long-term aging in a commercial finite element software with no need for extra subroutine codes or mesh superposition.
Ethylene propylene diene (EPDM) O-rings were aged in a compressed state at 75 °C, 100 °C, 125 °C and 150 °C for up to 183 days. Investigations of the experimental test results were used to identify material models and their parameters to develop a finite element model to simulate CS. The model was implemented in the finite element software ABAQUS/Standard® with a sequential temperature-displacement coupling.
Regarding the influence of temperature, the Arrhenius equation is adopted for the time-temperature relationship. The activation energy value that is required for the simulation is firstly determined from shifting the experimental CS results with the time-temperature superposition technique and plotting the shift factors in an Arrhenius diagram. The experiments were compared with the simulation results. Afterwards different activation energies were used in the simulation and discussed. A suitable choice of the activation energy value with regard to the reference temperature and the test temperature is presented. With the chosen activation energies, the match between numerical CS values after long-term aging and the experimental results was improved.
Recovery is an important measure for seal applications representing to which extent the elastomer regains its initial shape after deformation and release of an applied force. Compression set (CS) indicates the degree of recovery. Ethylene propylene diene rubber (EPDM) was aged at 75 ◦C, 100 ◦C, 125 ◦C and 150 ◦C for different ageing times up to five years and compression set measurements were performed at different times after disassembly and after additional tempering. Short- and long-term recovery up to one year after release for samples aged at 125 ◦C and 150 ◦C was also studied. To assess the curvature in the Arrhenius diagram that may occur due to non-sufficiently aged specimens, a degradation-rate based model was fitted to the CS data after tempering. For each ageing temperature, two decay fit functions were proposed, each with an activation energy and a corresponding degradative process. The influence of ageing on the leak-tightness after fast small partial release is investigated and estimated through the analysis of the shift factors from time temperature superposition (TTS) of CS measurements at different times after disassembly. Shift factors of CS measurement after 1 s and after additional tempering are in good agreement.