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In einem laufenden Kooperationsprojekt zwischen der BAM Berlin und der MPA Stuttgart wird das probabilistische Master Curve (MC) Konzept nach ASTM E1921 hinsichtlich seiner Übertragbarkeit auf ferritisches Gusseisen mit Kugelgraphit (DCI) untersucht. Zielsetzung ist die Bereitstellung einer Vorgehensweise zur Bestim-mung und Bewertung der dynamischen Bruchzähigkeit von DCI im Übergangsbe-reich. In diesem Beitrag wird der Versuchswerkstoff der Festigkeitsklasse GJS-400 näher vorgestellt. Ferner werden erste Ergebnisse von zwei bruchmechanischen Versuchsserien an SE(B)140-Großproben bei Belastungsraten von ca. 6∙104 MPa√ms-1 bei -40 °C und 60 °C präsentiert.
Within fracture mechanics safety assessment for steels in nuclear technology, the probabilistic master curve (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. However, for ferritic ductile cast iron (DCI), a systematic review of potential modifications to the assumptions and the procedure according to ASTM E1921 and an associated validation are still lacking. For this reason, the application of the fracture mechanics MC concept to ferritic ductile cast iron is being investigated in a joint research project between MPA Stuttgart and BAM Berlin. According to IAEA SSG-26, the safety assessment of DCI containers for transport as well as storage of radioactive materials is based on the fracture mechanics criterion of general crack initiation exclusion. Within this context, the focus of the research project is on brittle fracture. The goal is the establishment of a methodology to determine and assess dynamic fracture toughness values of DCI in the ductile-to-brittle transition regime, using samples extracted from a component.
The experimental program uses a DCI material of grade GJS-400, which is generally used for transport and storage containers of radioactive materials. Firstly, a basic mechanical-technological material characterization will be performed consisting of tensile, Charpy and Pellini tests. Following the material characterization, the loading rate relevant for brittle fracture will be determined using instrumented C(T)25-fracture tests at a temperature of -40°C. To be able to transfer the relevant loading rate to other specimen geometries or sizes, the time-dependent course of the Weibull stress is assessed via numerical analyses. With this information an extensive experimental program consisting of DC(T)9, C(T)25, C(T)50, SE(B)10, SE(B)25 and SE(B)140 specimens is performed. These experimental results will be further supported by numerical and fractographic analyses. The empirical experimental database developed for DCI materials will be used to assess the applicability of the MC concept for DCI materials under dynamic loading.
At the current time manufacturing and precracking of test specimens is ongoing. The authors provide first results concerning the mechanical-technological material characterization, the relevant loading rate for brittle fracture determined via C(T)25 tests, and some early results of other specimen types. These experimental results are also supported by preliminary numerical and statistical analyses.
The BAM-GGR 016 explains the requirements of ADR, RID, ADN and IMDG for management systems for design, manufacturing, testing, documentation, use, maintenance, and inspection of packagings for package designs for the transport of radioactive material not requiring competent authority approval for all designs of exempted packages, packages type IP-1, IP-2, IP-3 and type A.
Bei der Betonage von Abdichtbauwerken können Temperaturen bis 110 °C und hohe Drücke bis 5 MPa erreicht werden. Mittels Technikumsversuchen wurde nachgewiesen, dass Mikroakustik- und Ultraschallsensoren zum Einbau in Demonstrationsbauwerke geeignet sind.
Die Untersuchungen wurden bei der BAM (Bundesanstalt für Materialforschung und -prüfung) in Berlin durchgeführt. Es ist vorgesehen, die Sensoren in Morsleben am Demonstrationsbauwerk im Anhydrit einzusetzen.
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.
Ultrasonic echo testing is a well-established non-destructive testing technique used to investigate the interior of concrete structures in civil engineering. For the reconstruction of internal features, methods similar to seismic imaging are applied. To improve the ultrasonic imaging capabilities, we investigate the potential of a newly available 3D matrix mode measurement device in combination with advanced imaging techniques.
Commonly used ultrasonic array measurement devices rely on several ultrasonic transducers coupled to act as one transducer that transmits or receives transversal ultrasound waves. Although this approach is well suited to detect elongated structures such as pipes and rebars, it has limitations if 3D embedded objects are present. Hence, the ultrasonic device MiraA1040Pro used here is designed to measure the ultrasonic wavefield on a 4 by 16 matrix of individual transducers. For testing, we used laboratory specimens with linear and spheric embedded structures first. Both linear and matrix mode measurements were conducted and analysed using the Synthetic Aperture Focusing Technique
(SAFT) which is similar to Kirchhoff migration. Second, we used a data set from Teutschenthal mine at a shotcrete specimen constructed to simulate engineered barriers tailored for nuclear waste repositories. Ultrasonic measurements were investigated as a tool for quality assurance of these structures. The specimen contains both artificial 3D artifacts as well as naturally occurring defects such as a delamination. To make use of the advanced data acquisition, focusing 3D imaging techniques are applied to further improve the imaging quality.
Generally, the reconstructed images from the 3D matrix mode data have a lower level of Signal-to-Noise-Ratio than the 2D linear mode because the source signal is weaker compared to the linear mode. Moreover, complex wave conversions of the SH-transversal wave occur that are reduced in the 2D linear scenario. However, clearer images can be achieved at non-linear features, particularly with focusing imaging methods. Thus, combining 3D data acquisition techniques with advanced imaging methods improves the success in imaging complex concrete structures. This is of particular interest for thick concrete structures in nuclear barrier systems or foundations. Additionally, we demonstrate opportunities to use well-constrained test laboratory scenarios from non-destructive testing as a practical test case for geophysical methods.
Evaluation of Heat Fluxes in Fire Reference Test Conducted in BAM Propane Gas Fire Test Facility
(2024)
Packages for the transport of intermediate- and high-level radioactive waste are designed to withstand severe accidents. The International Atomic Energy Agency (IAEA) has established specific mechanical and thermal tests. Packages for the transport of radioactive material must withstand these tests to comply with the Regulations for the Safe Transport of Radioactive Materials IAEA [IAEA (2018)]. A fire reference package was developed with the primary objective to demonstrate that the fire meets the regulatory requirements. Another aim is to characterise the boundary conditions of the actual fire as input parameters for thermo-mechanical simulations. A simple method to characterise the boundary conditions of a real steady state fire with a fire reference package is presented. The thermal test mainly consists of a 30 minute fully engulfing 800°C pool fire or an equally severe fire, such as a propane gas fire. The fire reference tests are performed prior to the actual fire test with the real package. The heat fluxes into the package depend substantially on the fire characteristics and the surface temperature of the package. To investigate local and overall heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for repeated use. The fire reference package presented in this paper represents the outer geometry of a small transport container for radioactive material and is used as a device in civil engineering. It is designed as a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 182 mm and a diameter of 102 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material. Open-air fire tests were performed in a BAM propane gas fire test facility with the fire reference package. The measured temperatures are used to determine the changes of heat fluxes into the fire reference package in relation to the package surface temperature. The calculated heat fluxes allow its fitting to express the thermal exposure as simple mathematical boundary condition. Therefore, in a first approach, fire properties such as adiabatic surface temperature (AST) as proposed by Wickström et al. (2007), convection coefficient and emissivity are determined mathematically fitting the heat flux development presented in this paper. The evaluated results provide an initial picture of local fire characteristics of the conducted propane gas fire and are a further development of previous works from Feldkamp et al. (2020). The results can be used in thermal and thermo-mechanical models to simulate the load on the real transport package in fire. The test shows that the examined propane gas fire covers the IAEA-fire over a wide range of surface temperatures with the chosen test stand parameters.
Evaluation of Heat Fluxes in Fire Reference Test Conducted in BAM Propane Gas Fire Test Facility
(2024)
Packages for the transport of intermediate- and high-level radioactive waste are designed to withstand severe accidents. The International Atomic Energy Agency (IAEA) has established specific mechanical and thermal tests. Packages for the transport of radioactive material must withstand these tests to comply with the Regulations for the Safe Transport of Radioactive Materials IAEA [IAEA (2018)]. A fire reference package was developed with the primary objective to demonstrate that the fire meets the regulatory requirements. Another aim is to characterise the boundary conditions of the actual fire as input parameters for thermo-mechanical simulations. A simple method to characterise the boundary conditions of a real steady state fire with a fire reference package is presented. The thermal test mainly consists of a 30 minute fully engulfing 800°C pool fire or an equally severe fire, such as a propane gas fire. The fire reference tests are performed prior to the actual fire test with the real package. The heat fluxes into the package depend substantially on the fire characteristics and the surface temperature of the package. To investigate local and overall heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for repeated use. The fire reference package presented in this paper represents the outer geometry of a small transport container for radioactive material and is used as a device in civil engineering. It is designed as a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 182 mm and a diameter of 102 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material. Open-air fire tests were performed in a BAM propane gas fire test facility with the fire reference package. The measured temperatures are used to determine the changes of heat fluxes into the fire reference package in relation to the package surface temperature. The calculated heat fluxes allow its fitting to express the thermal exposure as simple mathematical boundary condition. Therefore, in a first approach, fire properties such as adiabatic surface temperature (AST) as proposed by Wickström et al. (2007), convection coefficient and emissivity are determined mathematically fitting the heat flux development presented in this paper. The evaluated results provide an initial picture of local fire characteristics of the conducted propane gas fire and are a further development of previous works from Feldkamp et al. (2020). The results can be used in thermal and thermo-mechanical models to simulate the load on the real transport package in fire. The test shows that the examined propane gas fire covers the IAEA-fire over a wide range of surface temperatures with the chosen test stand parameters.
Die BAM-GGR 016 erläutert die Anforderungen des ADR, RID, ADN und IMDG hinsichtlich notwendiger Maßnahmen zum Qualitäts- und Alterungsmanagement bei der Auslegung, Herstellung, Prüfung, Dokumentation, Gebrauch, Wartung und Inspektion von nicht-zulassungspflichtigen Versandstücken für radioaktive Stoffe für alle Bauarten von freigestellten Versandstücken, Industrieversandstücken des Typs IP-1, IP-2 und IP-3 und Versandstücken des Typs A.
Im laufenden Verbundprojekt MCGUSS untersuchen dieProjektpartner MPA Stuttgart und BAM Berlin das probabilistische Master Curve (MC) Konzept nach ASTM E1921 hinsichtlich seiner Übertragbarkeit auf ferritisches Gusseisen mit Kugelgraphit (DCI).
Ziel des Projektes ist die Entwicklung einer Methode zur Bestimmung und Bewertung der dynamischen Bruchzähigkeit von DCI im Übergangsbereich. In diesem Beitrag werden die Eigenschaften des verwendeten Werkstoffs der Festigkeitsklasse GJS-400 vorgestellt und die sich daraus ergebenden Probenentnahmestellen näher erläutert.
Weiterhin werden erste Ergebnisse der C(T)25-Versuchsserien vorgestellt, die bei Belastungsraten von ca. 5∙103 MPa√ms-1, 5∙104 MPa√ms-1 und 5∙105 MPa√ms-1 und einer Temperatur von -40 °C durchgeführt wurden.
In einem laufenden Kooperationsprojekt zwischen der BAM Berlin und der MPA Stuttgart wird das probabilistische Master Curve (MC) Konzept nach ASTM E1921 hinsichtlich seiner Übertragbarkeit auf ferritisches Gusseisen mit Kugelgraphit (DCI) untersucht.
Zielsetzung ist die Bereitstellung einer Vorgehensweise zur Bestimmung und Bewertung der dynamischen Bruchzähigkeit von DCI im Übergangsbereich. In diesem Beitrag wird der Versuchswerkstoff der Festigkeitsklasse GJS-400 näher vorgestellt.
Ferner werden erste Ergebnisse von zwei bruchmechanischen Versuchsserien an SE(B)140-Großproben bei Belastungsraten von ca. 6∙104 MPa√ms-1 bei -40 °C und -60 °C präsentiert.
The safety demonstration of heavy weight type B transport packages used for storing spent nuclear fuel (SNF) or radioactive waste is ensured by a combination of physical testing and numerical calculations. While experiments are performed in accordance with the IAEA regulations for selected drop scenarios, Finite-Element-Method (FEM) simulations are used to predict the most damaging case and to investigate additional drop positions. BAM as competent authority in Germany has performed different investigations of a welding seam for a typical large transport package made of A508 forged steel, where the bottom plate is welded to the cylindrical shell. The package has a mass of approx. 120 t. Results of physical drop tests with a full-scale model and accompanying preliminary FEM simulations are presented to determine the decisive stresses in the welding seam. A drop test only represents one set of a package and test parameters. A further parameter analysis is considered to account for allowable variations of packaging properties (e.g. resulting from the manufacturing process) and, based on IAEA requirements, the temperature dependence of the material behaviour. The results of the stress analyses from the drop test and the simulation form the basis and provide the input parameters for a fracture mechanics analysis. In addition to the IAEA specifications, further standards are taken into account for an in-depth investigation, see R6 [1], BS 7910 [2] and API 579-1/ASME FFS1 [3]. All the above-mentioned standards require a manufacturer-specific defect analysis with respect to size and position. Both result from the welding process and the following heat treatment regime. The maximum defect sizes are ensured with non-destructive test methods (such as ultrasonic or particle methods) as integral part of the manufacturing process of the welding seam. Another important parameter in the welding process is the residual stress (secondary stress). The combination of the primary and secondary stress determines the total stress in the welding seam. The most damaging case of the welding seam is determined and evaluated with help of the abovementioned standards and taking into account the IAEA requirements with respect to defect sizes, material properties, primary and residual stress, yield strength etc.
Experimental and numerical investigation of prestressed bolt connections under lateral displacements
(2022)
The containment system of transport packages for spent nuclear fuel and high-level waste usually includes bolted lids with metal gaskets. The packages are assessed to specific transport conditions which are specified in the IAEA safety standards SSR-6 (IAEA 2018). These transport conditions, especially the so-called accident conditions of transport, imply high dynamic loading on the lids and the bolt connections of the package. The response of the lid systems on the mechanical accident conditions is generally investigated by experimental drop tests or numerically, e.g., by finite element analyses. The interpretation of the drop test results for the verification of the numerical models is often not obvious due to the complex superposition of different effects in the real tests. BAM has started a research project to get a better understanding about the behavior of prestressed bolt connections under loadings typical for these drop tests. In this context an experimental test set-up was developed to investigate the response of a single bolt connection under a prescribed lateral displacement of clamped parts. The bolt is instrumented by strain gauges to get the pretensional, the torsional and the bending stress in the bolt shank. Furthermore, the lateral movement and the tilt of the bolt head is measured during the test. A finite element model of the test set-up has been created in Abaqus FEA (Simulia 2021). The very detailed instrumentation of the test set-up shall give the opportunity to investigate and validate the numerical model. The aim of this paper is to give an overview about the proposed research project and to present first results.
Within fracture mechanics safety assessment for steels in nuclear technology, the probabilistic master curve (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. However, for ductile cast iron (DCI), a systematic review of potential modifications to the assumptions and the procedure according to ASTM E1921 and an associated validation are still lacking. For this reason, the application of the fracture mechanics MC concept to ferritic ductile cast iron is being investigated in a joint research project between MPA Stuttgart and BAM Berlin. According to Hojo et al. (2008) the MC concept is already in use for the brittle fracture safety assessment of ferritic forged steel containers, which would suggest an extension to ductile cast iron containers. In compliance with IAEA SSG-26, the safety assessment of DCI containers for transport as well as storage of radioactive materials is based on the fracture mechanics criterion of general exclusion of crack initiation. Within this context, the focus of the current research project is on the further development of brittle fracture safety assessment methods. Therefore, a methodology shall be established to determine and assess dynamic fracture toughness values of DCI in the ductile-to-brittle transition regime, using only a low number of small fracture mechanics test specimens extracted from a component. For the experimental program, a DCI material of grade GJS-400 (GGG-40) has been produced as it is generally used for transport and storage containers for nuclear material. A basic mechanical-technological material characterization consisting of tensile, Charpy, and Pellini tests has been planned and 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 at various loading rates and temperatures. The specimen testing is divided between the project partners based on specimen geometry. MPA Stuttgart has been tasked with the testing of the C(T), DC(T) and Pellini specimens, while BAM Berlin was assigned the SE(B), Charpy and tensile tests. These tests allow for the determination of the loading rate relevant for brittle fracture via C(T)25 specimens and an assessment of size effects by transferring the relevant loading rate to other specimen geometries and sizes. Numerical analyses of different specimen geometries and test setups are planned to investigate the influence of stress triaxiality and loading rate. Furthermore, the time-dependent course of the Weibull stress is assessed via these numerical analyses. Throughout the experimental program, the microstructure-property relationship will also be investigated based on quantitative metallographic and fractographic analyses like scanning electron microscope images and metallographic sections. With the acquired experimental database fundamental assumptions of the MC concept, like the weakest link model and the Weibull distribution, will be assessed for the application to DCI. In the end, the applicability of the MC concept will be evaluated and if necessary, modifications to the MC concept for DCI will be proposed. This paper will present the preliminary results of the GJS400 material characterization and the related selection of test specimen extraction points. First results of the C(T)25 and SE(B)140 specimen testing will also be discussed supported by first metallographic analyses.
The Large Aperture Ultrasound System (LAUS) developed at BAM is known for its ability to penetrate thick objects, especially concrete structures commonly used in nuclear waste storage and other applications in civil engineering. Although the current system effectively penetrates up to ~9 m, further optimization is imperative to enhance the safety and integrity of disposal structures for radioactive or toxic waste. This study focuses on enhancing the system’s efficiency by optimizing the transducer spacing, ensuring that resolution is not compromised. An array of twelve horizontal shear wave transducers was used to find a balance between penetration depth and resolution. Systematic adjustments of the spacing between transmitter and receiver units were undertaken based on target depth ranges of known reflectors at depth ranges from 5 m to 10 m. The trade-offs between resolution and artifact generation were meticulously assessed. This comprehensive study employs a dual approach using both simulations and measurements to investigate the performance of transducer units spaced at 10 cm, 20 cm, 30 cm, and 40 cm. We found that for depths up to 5 m, a spacing of 10 cm for LAUS transducer units provided the best resolution as confirmed by both simulations and measurements. This optimal distance is particularly effective in achieving clear reflections and a satisfactory signal-to-noise ratio (SNR) in imaging scenarios with materials such as thick concrete structures. However, when targeting depths greater than 10 m, we recommend increasing the distance between the transducers to 20 cm. This increased spacing improves the SNR in comparison to other spacings, as seen in the simulation of a 10 m deep backwall. Our results emphasize the critical role of transducer spacing in achieving the desired SNR and resolution, especially in the context of depth imaging requirements for LAUS applications. In addition to the transducer spacing, different distances between individual sets of measurement positions were tested. Overall, keeping the minimal possible distance between measurement position offsets provides the best imaging results at greater depths. The proposed optimizations for the LAUS in this study are primarily relevant to applications on massive nuclear structures for nuclear waste management. This research highlights the need for better LAUS efficiency in applications such as sealing structures, laying the foundation for future technological advances in this field.
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.
Within fracture mechanics safety assessment for steels in nuclear technology, the probabilistic master curve (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. However, for ductile cast iron (DCI), a systematic review of potential modifications to the assumptions and the procedure according to ASTM E1921 and an associated validation are still lacking. For this reason, the application of the fracture mechanics MC concept to ferritic ductile cast iron is being investigated in a recently started joint research project between MPA Stuttgart and BAM Berlin. According to the IAEA guidelines SSG 26, the safety assessment of DCI containers for transport as well as 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. Therefore, a 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 will be 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 and 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. 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.
In dem Vortrag werden der Forschungsansatz, die Ziele, das Untersuchungsprogramm und die erwarteten Ergebnisse des BMUVForschungsvorhabens MCGUSS "Untersuchung des Master Curve-Konzeptes für ferritisches Gusseisen mit Kugelgraphit" vorgestellt. Mit Blick auf bruchmechanische Sicherheitsbewertungen in der Kerntechnik werden Stähle momentan mit dem deterministischen ASME Referenzkurvenkonzept, ergänzt durch das probabilistische Master Curve-Konzept nach ASTM E1921, bewertet. Das Master Curve-Konzept wird bereits für die Bewertung von ferritischen Schmiedestahlbehältern verwendet.
Bewertungsmethoden für ferritisches Gusseisen mit Kugelgraphit sind im Moment unvollständig. Normen wie die nationale BDG-Richtlinie P300 oder der internationale ASME-Code, Sect. III, Div. 3, WC-2330 decken Sprödbruch nicht ausreichend ab. Die Verwendung des Master Curve-Konzepts für ferritisches Gusseisen mit Kugelgraphit würde eine bruchmechanische Bewertung via einer KJc,d Verteilung erlauben und gleichzeitig Größeneffekte und einen Temperaturbereich von 100 K berücksichtigen.
In dem Poster werden der Forschungsansatz, die Ziele, das Untersuchungsprogramm und die erwarteten Ergebnisse des BMUV-Forschungsvorhabens MCGUSS "Untersuchung des Master Curve-Konzeptes für ferritisches Gusseisen mit Kugelgraphit" vorgestellt. Mit Blick auf bruchmechanische Sicherheitsbewertungen in der Kerntechnik werden Stähle momentan mit dem deterministischen ASME-Referenzkurvenkonzept, ergänzt durch das probabilistische Master Curve-Konzept nach ASTM E1921, bewertet. Das Master Curve-Konzept wird bereits für die Bewertung von ferritischen Schmiedestahlbehältern verwendet.
Bewertungsmethoden für ferritisches Gusseisen mit Kugelgraphit sind im Moment
unvollständig. Normen wie die nationale BDG-Richtlinie P300 oder der internationale ASME-Code, Sect. III, Div. 3, WC-2330 decken Sprödbruch nicht ausreichend ab. Die Verwendung des Master Curve-Konzepts für ferritisches Gusseisen mit Kugelgraphit würde eine bruchmechanische Bewertung via einer KJc,d Verteilung erlauben und gleichzeitig Größeneffekte und einen Temperaturbereich von 100 K berücksichtigen.
Within fracture mechanics safety assessment for steels in nuclear technology, the probabilistic master curve (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. However, for ductile cast iron (DCI), a systematic review of potential modifications to the assumptions and the procedure according to ASTM E1921 and an associated validation are still lacking. For this reason, the application of the fracture mechanics MC concept to ferritic ductile cast iron is being investigated in a recently started joint research project between MPA Stuttgart and BAM Berlin. According to the IAEA guidelines SSG 26, the safety assessment of DCI containers for transport as well as 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. Therefore, a 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 will be 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 and 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. 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.
The safe storage of radioactive material in nuclear waste repositories is a major task of our societies. The sealing function of such repositories is depending on the ability of the applied construction materials to form and maintain a geotechnical barrier with the surrounding rock. Corrosion of the construction material can lead to leakage of radioactive compounds and must therefore be avoided or minimized. Since concretes based on conventional Portland cement have not been found suitable in this context, alternative binders must be evaluated.
One such class of binders are hybrid cements, which are blends of low fractions of Portland cement or Portland clinker and high fractions of supplementary cementitious materials, such as blast furnace slag and fly ash, activated by an alkali salt. Besides a low heat of hydration and a sufficient early compressive strength [1], chemical similarities with ancient Roman concrete suggest an excellent durability in saline environments [2], which makes these cements potentially suitable for applications in nuclear waste repositories in evaporite rock, such as the Morsleben repository in Germany.
In the present study, two previously designed and characterized [3] hybrid cements, an alkali-activated slag/fly ash blend, and an OPC paste were studied regarding their resistance against corrosion in an aggressive saline solution. The saline solution was designed by the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) [Association for Facility and Reactor Safety] as a reference representing a solution forming as the result of contact of surface water with evaporite rock. The cement pastes were exposed to the saline solution up to 70 days and characterized by X-ray diffraction, thermogravimetric analysis and spatially resolved X-ray fluorescence spectroscopy. In addition, thermodynamic modelling was performed to simulate the alterations of the phase assemblage with increasing exposure to the saline solution and, thus, provide indications on the long-term durability of the cement pastes.
The experimental results revealed a correlation between the Portland clinker content of the cements and the resistance of the cement pastes against attack by the saline solution. This outcome was related to the formation of portlandite when sufficient clinker was available, which maintained the pore solution pH at ~12.5, and thus prevented the dissolution of cementitious phases. Once portlandite was consumed, C-N-A-S-H and ettringite dissolved and released calcium, aluminium, and hydroxide ions in the solution, maintaining the pH at ~10. In this pH range, the formation Cl-AFm phases was observed. At lower pH values, i.e., extended exposure durations, gypsum was the major corrosion product. The pH-dependent dissolution and formation of phases lead to pronounced zonation in the exposed cement pastes. Thermodynamic modelling indicated that the dissolved silicon from C-N-A-S-H reacts with magnesium ions in the saline solution to form M-S-H, and that long-term exposure eventually leads to a material rich in amorphous silica and brucite.
The presentation gives an overview about the current experiences of BAM regarding ageing management of transport packages in Germany.
The concept of the BAM-GGR 023 for ageing management of transport packagings will be described.
The fundamentals of BAM-GGR 023 will be shown, including the following key aspects:
-the organization of an Ageing Management System (AMS),
-the issue of an Ageing Management Plan (AMP),
-the ageing evaluation as a part of the AMP,
-the Ageing Surveillance Program (ASP) as a part of the AMP,
-the Ageing Management Documentation (AMD).
Die Regelwerksanforderungen für Verpackungen für den Transport radioaktiver Stoffe basieren auf den Empfehlungen der IAEA und werden über die verkehrsträgerspezifischen internationalen Regelwerke in nationales Recht umgesetzt.
Um die Anforderungen des Regelwerkes zu erläutern, entwickelt die Bundesanstalt für Materialforschung und -prüfung (BAM) Gefahrgutregeln (BAM-GGR), welche die Entwickler, Hersteller und Betreiber von Behältern über die von der BAM angewandten Verfahren bei der Umsetzung gesetzlicher Bestimmungen informieren und unterstützen sollen.
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.
Engineering barrier systems are a crucial part of the safe underground disposal of radioactive waste, particularly in salt as a host rock. Sealing structures made of tailored concrete are under test at various sites in Germany. To characterise the material properties of the concrete and potentially also the bonding to the host rock as part of the necessary subsurface structures, we apply non-destructive testing making use of advanced geophysical imaging methods.
Ultrasonic investigations have been carried out at underground concrete structures in the Teutschenthal mine in Germany. Here, we show results from two distinct constructions produced in a shotcrete procedure. Our goal is to analyse the performance of ultrasonics to be used as quality assurance for sealing structures made from shotcrete. First, a ~10 m long shotcrete structure is investigated with the help of a unique Large Aperture Ultrasonic System (LAUS) allowing for depth penetration of > 9 m. We perform measurements at the front and from the side of the construction. Second, we obtain results from a 1 m thick shotcrete body containing several artificial defects (width up to 8 cm). Ultrasonic testing data were acquired using a commercial multi-static device. Additionally, a new device measuring with 3D mode instead of line mode is applied and preliminary results will be shown.
Generally, the acquired ultrasonic data are analysed by the Synthetic Aperture Focusing Technique that is commonly applied in non-destructive testing. As a result, reflectors in the analysed shotcrete structures are imaged. Individual reflections from internal features and particularly the opposite wall are identified. An unexpected delamination wider than 30 cm is clearly imaged and later verified by boreholes. Thus, the method is – in general – suitable to serve as a quality measure. However, particularly the small and deep artificial defects can hardly be identified in the resulting images. Thus, we use advanced geophysical imaging methods to further enhance the quality of the obtained images. The recorded ultrasonic energy is focused to the physically reflective origin in the analysed volume. First results clearly show that we successfully improve the image quality regarding noise level and artifacts and hence facilitate the detection of objects.
In total, we present a valuable experiment under realistic conditions for underground sealing structures made from shotcrete, where the locations of artificial reflectors are partly known. This experiment serves as a unique basis to analyse the performance of advanced analysis methods to obtain high-quality images of the structure’s interior. Hence, the developed ultrasonic testing and analysis schemes can serve as a part of quality assurance that will help to enable safe sealing structures for nuclear waste disposal.
Acceptable limits for activity release from transport casks for high-level radioactive material specified in the IAEA regulations must be kept by the integrity of cask body and the cask sealing system. BAM as the German competent authority for mechanical, thermal and containment assessment of packages liable for approval verifies the activity release compliance with the regulatory limits. One of the fundamental aspects in assessment is the specification of conservative package design leakage rates.
To ensure the required package tightness for both, storage, and transport of the cask before and after storage usually metal seals of the Helicoflex® Type are used. Due to the long-term use the seal behavior is influenced by temperature and time. The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport specified in the regulations can have a significant effect on the leak tightness of the sealing system. Whereas the safety for application of new, non- aged Helicoflex® seals is verified sufficiently, there are still technical data gaps concerning the efficiency of aged Helicoflex® seals.
BAM performed experiments to learn more about the sealing efficiency of aged Helicoflex® seals with Aluminum and Silver outer jackets. The seals were compressed in test-flanges and for artificial ageing the complete flange systems were stored in an oven for several month at a high temperature. During the compression and decompression tests after the aging, load-deformation characteristics of the seals, and leakage rates were measured. With these tests a load situation was simulated, which can occur in the regulatory drop test of the cask: Under high impact loads the bolted lid can lift a little for a short moment, allowing a little movement of the seal, so that the contact area can change before compressing again.
The poster presentation will show details about test conditions and first results.
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.
BAM acts as authority and for service in safety assessment of packages for transport and storage of radioactive materials. We offer extensive test capabilities and application of analytical methods for design verification and simulation for all types of packages for the transport and storage of radioactive materials according with the international IAEA Regulations for the safe transport and for national storage acceptance criteria. BAM operates several test facilities for drop and stacking testing, leak testing and thermal testing. The large drop test tower allows dropping full-scale specimens up to 200,000 kg in any drop orientation as requested. The comprehensive test facilities combined with long-term experience, newest equipment and measurement devices according to the latest state-of-the-art technology ensures realisation of complex test campaigns for package safety evaluation.
Beyond that, non-destructive and destructive material test devices and experts are available. Equipment and application of all kinds of typical measurement categories can be offered for testing campaigns.
In recent years we performed testing of full-scale type B package models with complex handling and preparation procedures. The results were contributed for different package design approval procedures. Type A packages mainly designed for medical related transport purposes, were continuously tested according to the transport regulations over recent years as well. Moreover, we work on research topics with relevance to package safety. The mechanical behaviour of lid closure systems under transport and storage conditions and the thermal behaviour of impact limiters were recently of special importance for the assessment competencies of BAM and were investigated under use of our test facilities.
The paper describes the test facilities and capabilities for package design safety evaluation at BAM and shows examples from our recent work.
The consideration of ageing mechanisms is with integration into the IAEA-requirements now obligatory for the design of transport packages. For packages intended to be used for shipment after storage the consideration of the effects of ageing mechanisms during storage in safety analyses and the implementation of corresponding instructions for operation and maintenance are required. Therefore, these packages have been maintained during storage in a manner that all requirements specified in the IAEA-requirements and in the applicable certificates of approval have been fulfilled.
BAM is the competent authority in Germany for evaluating the thermal and mechanical design of packages for the transport of radioactive materials. In this context, BAM also evaluates the aging mechanisms and aging effects occurring during the long-term storage period of such transport packages. BAM has developed a guideline for the implementation of ageing assessment and of the measures for ageing management into the approval procedure based on IAEA-requirements.
The goal of the presented paper is to introduce this guideline and the general approach for ageing management requirements and to describe the application to a specific package design. The type and amount of the ageing management not only depend on the use of the packaging, but also on the accessibility of the components that are considered in the ageing assessment. The different operating phases of a packaging are considered for the ageing assessment, such as loading, interim storage or transport.
The implementation of measures for ageing management is divided into three levels – systemic measures, package design related measures and documentation. The systemic measures are attributed to the general management system and define the whole activities for organization of ageing management like structure, responsibilities, documentation, reports and evaluation. The package design related measures are defined in an ageing management plan (AMP). Therefore, an ageing surveillance program (ASP) and, if necessary, a gap analysis program has to be developed. The ageing management documentation (AMD) ensures the continuous documentation of the compliance of a specific package to the approved package design, comprising mainly records resulting from operation and surveillance.
BAM acts as authority and for service in safety assessment of packages for transport and storage of radioactive materials. We offer extensive test capabilities and application of analytical methods for design verification and simulation for all types of packages for the transport and storage of radioactive materials according with the international IAEA Regulations for the safe transport and for national storage acceptance criteria. BAM operates several test facilities for drop and stacking testing, leak testing and thermal testing. The large drop test tower allows dropping full-scale specimens up to 200,000 kg in any drop orientation as requested. The comprehensive test facilities combined with long-term experience, newest equipment and measurement devices according to the latest state-of-the-art technology ensures realisation of complex test campaigns for package safety evaluation.
Beyond that, non-destructive and destructive material test devices and experts are available. Equipment and application of all kinds of typical measurement categories can be offered for testing campaigns.
In recent years we performed testing of full-scale type B package models with complex handling and preparation procedures. The results were contributed for different package design approval procedures. Type A packages mainly designed for medical related transport purposes, were continuously tested according to the transport regulations over recent years as well. Moreover, we work on research topics with relevance to package safety. The mechanical behaviour of lid closure systems under transport and storage conditions and the thermal behaviour of impact limiters were recently of special importance for the assessment competencies of BAM and were investigated under use of our test facilities.
The paper describes the test facilities and capabilities for package design safety evaluation at BAM and shows examples from our recent work.
Im Rahmen des Seminars "Angewandte Forschungsthemen (SAFT)" am Institut für Geowissenschaften der Christian-Albrechts-Universität zu Kiel werden verschiedene geophysikalische Methoden vorgestellt, die bei der zerstörungsfreien Prüfung im Bauwesen zum Einsatz kommen. Neben einem Überblick der Arbeiten an der BAM im Allgemeinen und dem Fachbereich im Besonderen geht es vor allem um den Einsatz von Ultraschallmethoden. Drei wesentliche Schwerpunkte werden dabei beleuchtet:
1) (Geophysikalische) Antworten auf Herausforderungen bei Ultraschall-Anwendungen
2) Material research and multi-sensory monitoring for concrete sealing structures in rock salt unterground repositories (Projekt SealWasteSafe)
und 3) Untersuchung von Spritzbeton-Probekörpern unter Tage mit Ultraschall zur Qualitätssicherung von Verschlussbauwerken.
Insgesamt werden diverse Möglichkeiten für die Geophysik für Forschung und Studierende im Bereich der zerstörungsfreien Prüfung im Bauwesen aufgezeigt.
BAM-GGR 023 was published in June 2022
BAM-GGR 023 gives guidance to applicants regarding ageing management for competent authority approved package designs
Ageing management with evaluation of ageing mechanisms according to para 613A of IAEA SSR-6 (Rev. 1) and their effects are part of the general management system.
Essential items are: AMS (systemic-related), AMP (package design-related) and AMD (package-related)
There are different requirements for organization and extent of ageing management depending on type of use
There are also different requirements for ageing evaluation depending on type of use for packaging/package, classification of components and their accessibility and replaceability
Wood is widely used in impact limiters of Transport casks for radioactive material. Encapsulated by an outer and inner Steel structure, spruce wood is offen applied in layers of altemating direction. The friction at the interfaces between these layers is of crucial importance for the impact and energy absorption e.g., at an accidental impact of a cask against a hard component.
In order to get detailed Information for corresponding numerical calculations, in this study the friction coefficient for the combinations wood-wood and wood-steel was measured in the temperature ränge between -40 °C and 90 °C according to the relevant stress conditions for such casks. Results show decreasing friction with increasing temperature, ranging from 0.43 at -40 °C to 0.22 for 90 °C for wood-steel combinations and from 0.3 at -40 °C to 0.24 at 90 °C to for a wood-wood combination.
In accordance with the IAEA transport regulations, the design of special form radioactive material (SFRM) shall resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval includes besides the program for physical tests (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance, and inspection.
SFRM source design plus management system shall ensure, that every specimen of the approved design is able to survive the severe mechanical and thermal tests at any time of its SFRM-working life.
Due to the long-term use of SFRM designs in most cases, the assessment of the source ageing behavior is an important aspect in the approval procedure. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. Besides of radioactive content, corrosion is a main factor for possible SFRM design degradation.
Although the IAEA Advisory Material SSG-26 already implies an indication of the need for considering ageing mechanisms, suitable amendments in the regulatory requirements of SSR-6 should be introduced to make the approval procedure more transparent and help to reduce rounds of questions by the authority. A supplementary requirement for considering of ageing mechanisms could be a helpful contribution to an international harmonization of the approval procedure.
This paper will describe major influencing factors to be considered to assess the ageing behavior of a SFRM design and will identify the need for a regulatory specification of a SFRM-working life as basis for the assessment of the SFRM design regarding time-dependent weakening. A proposal for an explicit requirement for consideration of ageing mechanisms in safety assessment of SFRM, which should be considered in the ongoing SSR-6 revision cycle, will be explained.
In accordance with the IAEA transport regulations, the design of special form radioactive material (SFRM) shall resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval includes besides the program for physical tests (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance, and inspection.
SFRM source design plus management system shall ensure, that every specimen of the approved design is able to survive the severe mechanical and thermal tests at any time of its SFRM-working life.
Due to the long-term use of SFRM designs in most cases, the assessment of the source ageing behavior is an important aspect in the approval procedure. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. Besides of radioactive content, corrosion is a main factor for possible SFRM design degradation.
Although the IAEA Advisory Material SSG-26 already implies an indication of the need for considering ageing mechanisms, suitable amendments in the regulatory requirements of SSR-6 should be introduced to make the approval procedure more transparent and help to reduce rounds of questions by the authority. A supplementary requirement for considering of ageing mechanisms could be a helpful contribution to an international harmonization of the approval procedure.
This paper will describe major influencing factors to be considered to assess the ageing behavior of a SFRM design and will identify the need for a regulatory specification of a SFRM-working life as basis for the assessment of the SFRM design regarding time-dependent weakening. A proposal for an explicit requirement for consideration of ageing mechanisms in safety assessment of SFRM, which should be considered in the ongoing SSR-6 revision cycle, will be explained.
The consideration of ageing mechanisms is with integration of the new para 613A into IAEA SSR-6 (Rev. 1) now obligatory for the design of transport packages. In addition, para 809(f) requires for packages intended to be used for shipment after storage the consideration of the effects of ageing mechanisms during storage in safety analyses and the implementation of corresponding instructions for operation and maintenance. Para 503(e) requires that all packaging components and radioactive contents have been maintained during storage in a manner that all requirements specified in IAEA SSR-6 (Rev.1) and in the applicable certificates of approval have been fulfilled.
The evaluation of ageing mechanisms and their effects including monitoring are part of BAM’s authority assessment tasks related to the mechanical and thermal package design and quality assurance aspects. BAM has compiled a guideline for the implementation of ageing assessment and of the measures for ageing management of the approval procedure based on requirements of IAEA SSR-6 (Rev.1). The guideline is applicable only for packages requiring a competent authority approval.
The paper aims to describe the structure of the guideline and the general approach for ageing management requirements. The type and amount of measures for ageing management depend mainly on the use of the package and on the ageing effects for the component, which result from relevant ageing mechanisms during package operation time.
The implementation of measures for ageing management is divided into three levels – systemic measures, package design related measures and documentation. The systemic measures are attributed to the general management system and define the whole activities for organization of ageing management like structure, responsibilities, documentation, reports and evaluation. The package design related measures are defined in an ageing management plan (AMP). These measures shall ensure that the anticipated changes of the package design under consideration of ageing effects still complies with the design approval specification. Therefore, an ageing surveillance program (ASP) and, if necessary, a gap analysis program shall be developed. The ageing management documentation (AMD) ensures the continuous documentation of the compliance of a specific package to the approved package design, comprising mainly records resulting from operation and surveillance.
The decommissioning or refurbishment of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, and coolant pumps, to list the major contributors. These components or objects are large in size and mass, measuring up to approximately 6 meters in diameter, up to 20 meters in length, and weighing over 400 000 kg. In many situations, the components are transported off-site to a storage, disposal or recycling/treatment facility. Previously, many large objects had to be transported under special arrangement.
The decommissioning or refurbishment of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, and coolant pumps, to list the major contributors. These components or objects are large in size and mass, measuring up to approximately 6 meters in diameter, up to 20 meters in length, and weighing over 400 000 kg. In many situations, the components are transported off-site to a storage, disposal or recycling/treatment facility. Previously, many large objects had to be transported under special arrangement
Packages for the transport of radioactive materials shall fulfil the requirements of the IAEA regulations for the safe transport. The requirements define mechanical and thermal test conditions including criteria ensuring the package design’s ability to withstand severe accidents and provide a high level of technical safety. Different methods can be used for safety demonstration showing compliance with the regulations.
The central part of a safety demonstration which is presented in this paper was a comprehensive drop test program with a full-scale model of a transport package accompanied by pre- and post-test FE analyses. Using full-scale drop test models allow the benefit that similarity and scaling issues become a significant smaller issue, additional material investigations can be limited and analyses for transferring test results to the original package design are reduced. Additionally, experience for the future serial packaging manufacturing and handling procedures can be collected in a very early state of the design approval process. The pre-test finite element analyses derived and justified the drop test program consisting of several drop sequences with different drop orientations of the specimen. The performance and the results of the drop test sequences shows the manageability and the advantage e.g., in view of the direct availability of test results for the package licensing. On the other hand, the drop test performance shows the difficulties during handling and the need for additional equipment during preparation of the specimen.
The package presented was intended for the transport and storage of compacted radioactive waste from reprocessing of spent nuclear fuel assemblies - designed and applied for approval by the AGC consortium. The project ended in 2021. The package design was characterized by a cask body made of a forged thick stainless-steel shell, a bolted double lid system with metallic gaskets and wood filled shock absorbers at both ends. The total mass of the entire transport package including content was 120,000 kg, the total length was about 7000 mm and the diameter approximately
3000 mm, both measures include the shock absorbers.
The paper provides an insight into the performance of a full-scale drop testing campaign within the package safety evaluation and shows some selected test results.
Packages for the transport of radioactive materials shall fulfil the requirements of the IAEA regulations for the safe transport. The requirements define mechanical and thermal test conditions including criteria ensuring the package design’s ability to withstand severe accidents and provide a high level of technical safety. Different methods can be used for safety demonstration showing compliance with the regulations.
The central part of a safety demonstration which is presented in this paper was a comprehensive drop test program with a full-scale model of a transport package accompanied by pre- and post-test FE analyses. Using full-scale drop test models allow the benefit that similarity and scaling issues become a significant smaller issue, additional material investigations can be limited and analyses for transferring test results to the original package design are reduced. Additionally, experience for the future serial packaging manufacturing and handling procedures can be collected in a very early state of the design approval process. The pre-test finite element analyses derived and justified the drop test program consisting of several drop sequences with different drop orientations of the specimen. The performance and the results of the drop test sequences shows the manageability and the advantage e.g., in view of the direct availability of test results for the package licensing. On the other hand, the drop test performance shows the difficulties during handling and the need for additional equipment during preparation of the specimen.
The package presented was intended for the transport and storage of compacted radioactive waste from reprocessing of spent nuclear fuel assemblies - designed and applied for approval by the AGC consortium. The project ended in 2021. The package design was characterized by a cask body made of a forged thick stainless-steel shell, a bolted double lid system with metallic gaskets and wood filled shock absorbers at both ends. The total mass of the entire transport package including content was 120,000 kg, the total length was about 7000 mm and the diameter approximately
3000 mm, both measures include the shock absorbers.
The paper provides an insight into the performance of a full-scale drop testing campaign within the package safety evaluation and shows some selected test results.
Fire Influence to Wood Filled Impact Limiters - Implications for the Package Design Safety Case
(2023)
Impact limiters with wooden components are widely used in the design of packages for transportation of radioactive material. In most designs, the wood is encapsulated with steel sheets. Impact limiters mainly determine the mechanical and thermal behaviour of the package in accident conditions of transport in accordance with the IAEA Regulations.
In context with research and development for package design approval competence, the thermal behaviour of heavy-weight packages was investigated at BAM with an artificially pre-damaged generic impact limiter design. Within this first investigation, the pre-damaged impact limiter with a diameter of 2.3 meters was mounted on a water tank simulating the thermal capacity of a cask during the fire test. The water tank is part of a water circulating system built of several components such as pump, heater, cooler, sliding valve, flow meter, thermocouples and control unit in order to measure the heat flux.
Furthermore, the investigations focus on the effects this additional heat generation would have on the cask and especially on the lid-closure system with the gasket.
The results of these experiments could find consideration in the safety case of the transport packages of radioactive material using wood filled impact limiters.
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.
Most transports of radioactive materials are carried out with packages not requiring competent authority approval of design. These encompass – in accordance with the IAEA SSR-6 [1] regulations – packages of the classification excepted, Industrial packages Type 1, 2 and 3 and Type A packages. Currently an upsurge in number and variation of these package designs can be seen in Germany, resulting from the phase out of nuclear energy in Germany as well as e. g. increased use of radioactive material for medical purposes.
A design assessment regarding the package safety is required in the international IAEA SSR-6 [1] regulations. BAM operates facilities for the performance of all regulatory tests required such as drop towers for a wide range of package masses and dimensions, fire test, leak tightness measurements and pressure test facilities. Experiences with several package types are shown.
Additionally, IAEA SSR-6 [1] requires the establishment of a management system for design, manufacture, maintenance, and repair of the packaging as well as for the preparation, consigning, loading, carriage, unloading and receipt of the package. Relevant for Germany, BAM has published guidance material on the process of management system acceptance in the technical guide BAM-GGR 016 [2]. The requirements encompass quality management plans for the manufacturing of packages including independent manufacturing surveillance and specific instructions for operation, maintenance, and repair of packagings. Examples for management system specifics and requirements are given.
Since the latest edition of the IAEA SSR-6 [1] regulations an ageing evaluation including systematic ageing management measures are required for all kind of package types. BAM is going to update the guidance material BAM-GGR 016 [2] to support the stakeholders with relevant information to fulfil the ageing aspect for packages not requiring competent authority approval. The paper explains how the ageing aspect may be included in the safety evaluation process and the management system measures and will give an outlook for the future guidance material.
Most transports of radioactive materials are carried out with packages not requiring competent authority approval of design. These encompass – in accordance with the IAEA SSR-6 regulations – packages of the classification excepted, Industrial packages Type 1, 2 and 3 and Type A packages. Currently an upsurge in number and variation of these package designs can be seen in Germany, resulting from the phase out of nuclear energy in Germany as well as e. g. increased use of radioactive material for medical purposes.
A design assessment regarding the package safety is required in the international IAEA SSR-6 regulations. BAM operates facilities for the performance of all regulatory tests required such as drop towers for a wide range of package masses and dimensions, fire test, leak tightness measurements and pressure test facilities. Experiences with several package types are shown.
Additionally, IAEA SSR-6 requires the establishment of a management system for design, manufacture, maintenance, and repair of the packaging as well as for the preparation, consigning, loading, carriage, unloading and receipt of the package. Relevant for Germany, BAM has published guidance material on the process of management system acceptance in the technical guide BAM-GGR 016. The requirements encompass quality management plans for the manufacturing of packages including independent manufacturing surveillance and specific instructions for operation, maintenance, and repair of packagings. Examples for management system specifics and requirements are given.
Since the latest edition of the IAEA SSR-6 regulations an ageing evaluation including systematic ageing management measures are required for all kind of package types. BAM is going to update the guidance material BAM-GGR 016 to support the stakeholders with relevant information to fulfil the ageing aspect for packages not requiring competent authority approval. The paper explains how the ageing aspect may be included in the safety evaluation process and the management system measures and will give an outlook for the future guidance material.
Friction coefficients for wood-wood and wood-steel interfaces in impact limiters for transport casks
(2023)
Wood is widely used in impact limiters of transport casks for radioactive material. Encapsulated by an outer and inner steel structure, spruce wood is often applied in layers of alternating direction. The friction at the interfaces between these layers is of crucial importance for the impact and energy absorption e.g., at an accidental impact of a cask against a hard target. In order to get detailed information for corresponding numerical calculations, in this study the friction coefficient for the combinations wood-wood and wood-steel was measured in the temperature range between -40°C and 90°C according to the relevant stress conditions for such casks. Results show decreasing friction with increasing temperature, ranging from 0.43 at -40°C to 0.22 for 90°C for wood-steel combinations and from 0.3 at -40°C to 0.24 at 90°C to for a wood-wood combination.
The safety demonstration of heavy weight type B transport packages used for storing spent nuclear fuel (SNF) or radioactive waste is ensured by a combination of physical testing and numerical calculations. While experiments are performed in accordance with the IAEA regulations for selected drop scenarios, Finite-Element-Method (FEM) simulations are used to predict the most damaging case and to investigate additional drop positions.
BAM as competent authority in Germany has performed different investigations of a welding seam for a typical large transport package made of A508 forged steel, where the bottom plate is welded to the cylindrical shell. The package has a mass of approx. 120 t. Results of physical drop tests with a full-scale model and accompanying preliminary FEM simulations are presented to determine the decisive stresses in the welding seam. A drop test only represents one set of a package and test parameters. A further parameter analysis is considered to account for allowable variations of packaging properties (e.g. resulting from the manufacturing process) and, based on IAEA requirements, the temperature dependence of the material behaviour. The results of the stress analyses from the drop test and the simulation form the basis and provide the input parameters for a fracture mechanics analysis. In addition to the IAEA specifications, further standards are taken into account for an in-depth investigation, see R6 [1], BS 7910 [2] and API 579-1/ASME FFS1 [3].
All the above-mentioned standards require a manufacturer-specific defect analysis with respect to size and position. Both result from the welding process and the following heat treatment regime. The maximum defect sizes are ensured with non-destructive test methods (such as ultrasonic or particle methods) as integral part of the manufacturing process of the welding seam. Another important parameter in the welding process is the residual stress (secondary stress). The combination of the primary and secondary stress determines the total stress in the welding seam. The most damaging case of the welding seam is determined and evaluated with help of the above-mentioned standards and taking into account the IAEA requirements with respect to defect sizes, material properties, primary and residual stress, yield strength etc.
The safety demonstration of heavy weight type B transport packages used for storing spent nuclear fuel (SNF) or radioactive waste is ensured by a combination of physical testing and numerical calculations. While experiments are performed in accordance with the IAEA regulations for selected drop scenarios, Finite-Element-Method (FEM) simulations are used to predict the most damaging case and to investigate additional drop positions.
BAM as competent authority in Germany has performed different investigations of a welding seam for a typical large transport package made of A508 forged steel, where the bottom plate is welded to the cylindrical shell. The package has a mass of approx. 120 t. Results of physical drop tests with a full-scale model and accompanying preliminary FEM simulations are presented to determine the decisive stresses in the welding seam. A drop test only represents one set of a package and test parameters. A further parameter analysis is considered to account for allowable variations of packaging properties (e.g. resulting from the manufacturing process) and, based on IAEA requirements, the temperature dependence of the material behaviour. The results of the stress analyses from the drop test and the simulation form the basis and provide the input parameters for a fracture mechanics analysis. In addition to the IAEA specifications, further standards are taken into account for an in-depth investigation, see R6, BS 7910 and API 579-1/ASME FFS1.
All the above-mentioned standards require a manufacturer-specific defect analysis with respect to size and position. Both result from the welding process and the following heat treatment regime. The maximum defect sizes are ensured with non-destructive test methods (such as ultrasonic or particle methods) as integral part of the manufacturing process of the welding seam. Another important parameter in the welding process is the residual stress (secondary stress). The combination of the primary and secondary stress determines the total stress in the welding seam. The most damaging case of the welding seam is determined and evaluated with help of the above-mentioned standards and taking into account the IAEA requirements with respect to defect sizes, material properties, primary and residual stress, yield strength etc.
Friction coefficients for wood-wood and wood-steel interfaces in impact limiters for transport casks
(2023)
Wood is widely used in impact limiters of transport casks for radioactive material. Encapsulated by an outer and inner steel structure, spruce wood is often applied in layers of alternating direction. The friction at the interfaces between these layers is of crucial importance for the impact and energy absorption e.g., at an accidental impact of a cask against a hard target. In order to get detailed information for corresponding numerical calculations, in this study the friction coefficient for the combinations wood-wood and wood-steel was measured in the temperature range between -40°C and 90°C according to the relevant stress conditions for such casks. Results show decreasing friction with increasing temperature, ranging from 0.43 at -40°C to 0.22 for 90°C for wood-steel combinations and from 0.3 at -40°C to 0.24 at 90°C to for a wood-wood combination.
Two hybrid alkaline cements (HAC) based on Portland clinker, ground granulated blast furnace slag (GGBFS), fly ash and sodium sulfate, as well as an alkali-activated GGBFS/fly ash blend and a Portland cement paste were exposed to a saturated saline solution for 70 days. The combined chemical attack of chloride, magnesium and sulfate ions and the associated changes of the phase assemblage of the materials were studied by X-ray diffraction, thermal analysis and spatially resolved X-ray fluorescence spectroscopy. The experimental results revealed dissolution of ettringite, C-N-A-S-H and calcite, and the formation of gypsum, Kuzel's salt and Friedel's salt; thermodynamic modeling indicated the formation of M-S-H. The resistance of the HAC against attack by the saline solution increased with Portland clinker fraction. The capacity of portlandite to maintain pH at values above 10 is found to be a major factor controlling the resistance of HAC against corrosion in the saline solution.
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.
Within the project SealWasteSafe, we advance construction materials and non-destructive monitoring concepts of sealing structures applied for underground disposal of nuclear waste. As these engineered barriers have high demands regarding structural integrity, an innovative alkali-activated material (AAM) that is highly suitable for the application in salt as a host rock is improved and tested on two laboratory scales. This AAM has a low heat evolution due to the reaction mechanism in comparison to common salt concretes based on Portland cement or magnesium oxychloride binders. Hence, crack formation due to thermally induced stress during the hardening process is reduced.
After successful laboratory tests with small specimens (height ~5 cm), comparably manufactured large cubic (edge length 70 cm) and cylindrical specimens (height 120 cm, diameter 40 cm) are equipped with sensing technologies to demonstrate the sensors´ technical capabilities. A comprehensive multi-sensory monitoring scheme is developed and investigated to characterize and compare the different material behaviour during the setting and hardening process of two materials: (1) the newly developed AAM-based mortars with salt aggregate, and (2) a blended Portland cement-based salt concrete as reference. The analysed parameters include temperature and humidity of the material, acoustic emissions, and strain variations recorded by fiber optic cables. Passive sensor systems based on radiofrequency identification technology (RFID) embedded in the concrete provide an interface for the wireless readout of various sensors. In parallel to the embedded RFID sensors, conventional cabled systems to read out the temperature and humidity measurements are installed for comparison. Additionally, a detailed inspection of the two large cubic specimens after a monitoring period of more than six months has been undertaken. Active thermography and ultrasonic echo measurements are used to reveal potentially occurring inner cracks from the surface. To verify the non-invasive results, a core sample (diameter 2 cm) was extracted from each of the investigated cubic specimens and analysed in detail with X-ray computed tomography.
Furthermore, ultrasonic methods are used for quality assurance to detect obstacles, cracks, and delamination at in-situ scale sealing structures. Experimental layout and applied imaging techniques are optimised to enhance the image quality for measurements from the front side of the engineered barrier. To characterize the inside of the test sealing structure and to improve the detection of potentially existing cracks, an ultrasonic borehole probe using the phased array technique is developed. First analyses at a half-spherical specimen coincide with modelling results and prove the reliability of the directional response caused by the phased array technique of the newly constructed ultrasonic borehole probe. Overall, the project SealWasteSafe helps to characterize construction materials and improves multi-sensory monitoring concepts and ultrasonic equipment for the sake of quality assurance. Particularly for salt as a host rock, this will help to design safe sealing structures for nuclear waste disposal.
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.
A small cylindrical fire reference package was designed. A fire test setup was created using the fire reference package in accordance with IAEA standards. Four propane gas fire tests were conducted to determine heat fluxes into the fire reference package. The initial test setup was modified to create the final design of the fire test facility. This allowed for the heat flux to be adjusted to meet the regulatory IAEA fire qualification criteria. Furthermore, a numerical model of the fire reference package was created using boundary conditions derived from the experimental data. The simulation results demonstrated good agreement with the experimental data and provided additional insights.
SFC is a work package in Eurad that investigates issues related to the properties of the spent nuclear fuel in the back-end of the nuclear fuel cycle. Decay heat, nuclide inventory, and fuel integrity (mechanical and otherwise), and not least the related uncertainties, are among the primary focal points of SFC. These have very significant importance for the safety and operational aspect of the back-end. One consequence is the operation economy of the back-end, where deeper understanding and quantification allow for significant optimization, meaning that significant parts of the costs can be reduced. In this paper, SFC is described, and examples of results are presented at about half-time of the work package, which will finish in 2024. The DisCo project started in 2017 and finished in November 2021 and was funded under the Horizon 2020 Euratom program. It investigated if the properties of modern fuel types, namely doped fuel, and MOX, cause any significant difference in the dissolution behavior of the fuel matrix compared with standard fuels. Spent nuclear fuel experiments were complemented with studies on model materials as well as the development of models describing the solid state, the dissolution process, and reactive transport in the near field. This research has improved the understanding of processes occurring at the interface between spent nuclear fuel and aqueous solution, such as redox reactions. Overall, the results show that from a long-term fuel matrix dissolution point of view, there is no significant difference between MOX fuel, Cr+Al-doped fuel, and standard fuels.
Characterization is a very important step in dealing with materials and waste streams generated during the operational and decommissioning phases of nuclear installations, including nuclear power plants. Characterization allows differentiation between materials that can be released from regulatory control and those that require further treatment and conditioning to become a stable waste form suitable for future storage and final disposal according to its waste classification. Characterization is also needed in the pre-disposal stages of radioactive waste management to demonstrate compliance with the waste acceptance criteria of the facilities that will accept the different waste forms. This work will present the strategies developed and implemented by the three projects for in-depth and accurate waste characterization and investigation of the different radioactive waste packages considered. CHANCE, MICADO, and PREDIS will present their goals, the methods developed, the technologies used and the (preliminary) results contributing to the improvement of the safety and the data and information quality of the waste packages analyzed at the different stages of the waste management process. Special emphasis will also be given to complementary approaches highlighting the usability of the technologies, the accessibility of the data, and the problem-solving of the three projects within the European panorama.
Für die sichere Endlagerung nuklearer Abfälle sind im Salzgestein Verschlussbauwerke aus Salzbeton als geotechnischen Barrieren vorgesehen. Um die Integrität der Bauwerke im Rahmen einer Qualitätssicherung nachzuweisen, eignet sich eine zerstörungsfreie Prüfung (ZfP) mit Ultraschall-Verfahren. Potenziell auftretende Risse oder Ablösungen sowie Einbauteile können detektiert werden. Um die Möglichkeiten und Grenzen der Ultraschall-Echo-Messungen sowie der Auswertemethoden zu untersuchen, wurden in einem aus Spritzbeton hergestellten Testbauwerk in der Grube Teutschenthal (bei Halle, Deutschland) verschiedene Störkörper während der Betonage eingebettet. Die Störkörper (Breite 3 cm bis ~8 cm) liegen in Tiefen von etwa 30 cm bis 80 cm und repräsentieren Hohlräume, Risse oder Inhomogenitäten des Betons. Von der Oberfläche führten wir Ultraschall-Echo-Messungen auf einer Fläche von etwa 100x150 cm² in einem 3D Modus durch. Die Auswertung erfolgte zunächst mithilfe der in der ZfP etablierten Synthetic-Aperture-Focusing Technique (SAFT), die der Kirchhoff Migration ähnelt. Insgesamt wird eine Vielzahl von Reflektoren im Bauwerk abgebildet, die jedoch nur in wenigen Fällen den eingebauten Störkörpern zugeordnet werden konnten, so dass auch die Grenzen des Verfahrens bei dieser Bauweise deutlich werden. Allerdings konnte eine ungeplante großflächige (>30 cm) Ablösung nachgewiesen und durch anschließende Bohrungen verifiziert werden. Um das strukturelle Abbild des Bauwerksinneren trotz des stark streuenden Spritzbetons zu verbessern, wenden wir fokussierende Migrationsverfahren an. Erste Ergebnisse zeigen, dass eine klarere Abbildung einzelner Reflektoren erreicht werden kann. Dies erleichtert die Detektion einzelner Objekte. Das hier präsentierte Experiment mit bekannten Reflektoren unter realen Bedingungen mit Spritzbeton unter Tage bietet hervorragende Möglichkeiten zur Überprüfung von methodischen Weiterentwicklungen im Bereich der Ultraschallanalyse. Durch die Verbesserung der Auswertemethodik und der darauf basierenden ultraschallbasierten Qualitätssicherungskonzepte werden insgesamt sicherere Verschlussbauwerke für nukleare Endlager ermöglicht.
Elastomer seals are mounted as barrier seals in lid systems of containers designed for transport and disposal of negligible heat generating radioactive waste and as auxiliary seals in spent fuel storage and transportation casks (dual purpose casks (DPC)). When the behavior of mounted seals under normal and hypothetical accident conditions of disposal and transport is to be simulated, a comprehensive knowledge of their complex mechanical properties at every state of aging is necessary. In previous works, BAM’s efforts in experimental investigations on specimen artificially aged at different temperatures and times and the implementation of the found results in finite element material models were presented. Additionally, our approaches to reproduce the aging process itself and to extrapolate the results of artificially accelerated aging to longer times were presented. Numerical simulations have shown that the behavior of the seal during mounting and one-sided pressurizing and the resulting performance values such as leakage rate strongly depend on the coefficient of friction (COF) between flange and seal. The friction coefficient, in turn, depends on the aging state of the elastomer material as several publications suggest (see below).
Dynamic COF between an exemplary ethylene propylene diene rubber (EPDM) material and a stainless steel ball were determined by using a self-designed linear oscillation tribometer. Unaged and artificially aged EPDM specimen stored for 30 days and 100 days at a temperature of 150 °C were tested. A stainless steel ball (d=10 mm) is brought in contact with the specimen’s surface und loaded by normal forces of 2.5 N, 5 N, 10 N and 20 N. During a reciprocating movement of the EPDM sheet, the horizontal force/friction force is continuously measured, and the COF can be derived.
It is well known that friction is a complex phenomenon especially in soft materials. It cannot be excluded that the measured friction force is influenced by additional force components, resulting from the ball’s grooving through the elastomer’s surface. This force depends on the penetration depth of the ball and on the resistance of the elastomer in its different states of aging. The latter results from microstructural changes i.e., chain scission and additional crosslinking that occur during aging which in turn influence the softening or hardening of the material.
A finite element (FE) ABAQUS® model was developed to reproduce the measurement process. It should help to better understand the physical mechanisms and to quantify the percentage of measured forces resulting from real friction on the one hand and forces resulting from unintended side effects that could falsify the result on the other hand. The behavior of the elastomer in its different states of aging is reproduced by a FE material model already presented in previous works of BAM.
Elastomer seals are mounted as barrier seals in lid systems of containers designed for transport and disposal of negligible heat generating radioactive waste and as auxiliary seals in spent fuel storage and transportation casks (dual purpose casks (DPC)). When the behavior of mounted seals under normal and hypothetical accident conditions of disposal and transport is to be simulated, a comprehensive knowledge of their complex mechanical properties at every state of aging is necessary. In previous works, BAM’s efforts in experimental investigations on specimen artificially aged at different temperatures and times and the implementation of the found results in finite element material models were presented. Additionally, our approaches to reproduce the aging process itself and to extrapolate the results of artificially accelerated aging to longer times were presented. Numerical simulations have shown that the behavior of the seal during mounting and one-sided pressurizing and the resulting performance values such as leakage rate strongly depend on the coefficient of friction (COF) between flange and seal. The friction coefficient, in turn, depends on the aging state of the elastomer material as several publications suggest (see below).
Dynamic COF between an exemplary ethylene propylene diene rubber (EPDM) material and a stainless steel ball were determined by using a self-designed linear oscillation tribometer. Unaged and artificially aged EPDM specimen stored for 30 days and 100 days at a temperature of 150 °C were tested. A stainless steel ball (d=10 mm) is brought in contact with the specimen’s surface und loaded by normal forces of 2.5 N, 5 N, 10 N and 20 N. During a reciprocating movement of the EPDM sheet, the horizontal force/friction force is continuously measured, and the COF can be derived.
It is well known that friction is a complex phenomenon especially in soft materials. It cannot be excluded that the measured friction force is influenced by additional force components, resulting from the ball’s grooving through the elastomer’s surface. This force depends on the penetration depth of the ball and on the resistance of the elastomer in its different states of aging. The latter results from microstructural changes i.e., chain scission and additional crosslinking that occur during aging which in turn influence the softening or hardening of the material.
A finite element (FE) ABAQUS® model was developed to reproduce the measurement process. It should help to better understand the physical mechanisms and to quantify the percentage of measured forces resulting from real friction on the one hand and forces resulting from unintended side effects that could falsify the result on the other hand. The behavior of the elastomer in its different states of aging is reproduced by a FE material model already presented in previous works of BAM.
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 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 measurements across the section of O-rings. Oxidation profiles were determined by applying the model of Wise et al. to the IRHD experimental results.
Suitable containers for receiving high-level radioactive waste are a key elements of every repository system concerning different host rocks like rock salt, claystone and crystalline rock as being considered in Germany. The disposal container design is decisive regarding boundary conditions for the transport, emplacement techniques, and the operational and long-term safety assessment. The systematic derivation of the requirements to be placed on disposal containers forms the basis for a targeted, comprehensible and transparent development of containers that meet the requirements. The presentation summarizes the outcomes of the research project KoBrA.
Data science (digitalisation and artificial intelligence) became more than an important facilitator for many domains in fundamental and applied sciences as well as industry and is disrupting the way of research already to a large extent. Originally, data sciences were viewed to be well-suited, especially, for data-intensive applications such as image processing, pattern recognition, etc. In the recent past, particularly, data-driven and physics-inspired machine learning methods have been developed to an extent that they accelerate numerical simulations and became directly usable for applications related to the nuclear waste management cycle. In addition to process-based approaches for creating surrogate models, other disciplines such as virtual reality methods and high-performance computing are leveraging the potential of data sciences more and more. The present challenge is utilising the best models, input data and monitoring information to integrate multi-chemical-physical, coupled processes, multi-scale and probabilistic simulations in Digital Twins (DTw) able to mirror or predict the performance of its corresponding physical twins. Therefore, the main target of the Topical Collection is exploring how the development of DTw can benefit the development of safe, efficient solutions for the pre-disposal and disposal of radioactive waste. A particular challenge for DTw in radioactive waste management is the combination of concepts from geological modelling and underground construction which will be addressed by linking structural and multi-physics/chemistry process models to building or tunnel information models. As for technical systems, engineered structures a variety of DTw approaches already exist, the development of DTw concepts for geological systems poses a particular challenge when taking the complexities (structures and processes) and uncertainties at extremely varying time and spatial scales of subsurface environments into account.
Multi-sensory monitoring and ultrasound for quality assurance at underground sealing structures
(2023)
Within the safety concepts of underground disposal sites of nuclear waste, engineered barriers play an important role. As these sealing structures have high demands concerning integrity, we aim at advancing the available construction materials, monitoring, and inspection techniques within the project SealWasteSafe. A specifically developed alkali-activated material is compared to classical salt concrete. A comprehensive multi-sensory monitoring scheme is used at 150-340 l specimens to monitor setting and hardening of both materials. All sensors are demonstrated to resist the highly alkaline environments. Besides cabled and wireless temperature and humidity of the materials, strain variations using fibre optic sensors and acoustic emissions are recorded over periods of at least 28 days, partly for more than eight months. After hardening of the specimens, further nondestructive evaluations using ultrasonic echo and thermographic measurements are conducted.
Preliminary results proof the suitability of the tested sensors and clearly highlight differences between the tested materials. Particularly, the newly developed alkali-activated material shows lower acoustic emission activity indicating less cracking activity. Additionally, unique ultrasonic methods will enable better images of potential internal objects and cracks at in-situ sealing structures. A largescale ultrasonic system is optimised to reliably detect objects at a depth exceeding 9 m while still obtaining a good resolution. Modelling studies show the potential of further increasing the distance between individual transducer arrays. Additionally, a new ultrasonic borehole probe using phased arrays allowing for beam focussing is constructed and tested. Laboratory measurements at a halfcylindrical concrete specimen coincide well with the previous modelling. In total, the presented safe materials, detailed monitoring approaches and ultrasonic quality assurance methods will help to obtain safe sealing structures within salt as a host rock. The concepts can partly be transferred to sealing structures in alternative host rocks and will also be valuable for non-nuclear waste repositories.
Multi-sensory monitoring and ultrasound for quality assurance at underground sealing structures
(2023)
Within the safety concepts of underground disposal sites of nuclear waste, engineered barriers play an important role. As these sealing structures have high demands concerning integrity, we aim at advancing the available construction materials, monitoring and inspection techniques within the project SealWasteSafe.
After successful laboratory analysis on the cm-scale, cylindrical and cuboid specimens on the m-scale are produced from classical salt concrete and a specifically developed alkali-activated material. A comprehensive multi-sensory monitoring scheme is applied to compare the setting process of both materials and to demonstrate the sensors’ resistance to highly alkaline environments. Besides temperature and humidity of the materials, strain variations using fibre optic sensors and acoustic emissions are recorded over periods of at least 28 days, partly for more than 8 months. Passive sensor systems based on radiofrequency identification technology (RFID) embedded in the concrete allow for wireless access to temperature and humidity measurements and are compared to conventional cabled systems. After hardening and removal of the outer casings of the specimens, further non-destructive evaluations using ultrasonic echo and thermographic measurements are conducted. Preliminary results clearly highlight differences between the tested materials, particularly showing lower acoustic emission activity for the newly developed alkali-activated material potentially indicating less phase changes or cracks.
Complementary, ultrasonic methods are improved to be used for quality assurance to detect obstacles, potential cracks and delamination at in-situ sealing structure scale. A unique large aperture ultrasonic system (LAUS) with depth penetration as large as 9 m has already successfully been applied at the test site in Morsleben, Germany, of the federal company for radioactive waste disposal (BGE). Modelling studies help to further optimize the measurement layout. Advanced imaging techniques applied to the modelled and measured data will further improve the obtained images of internal structures. Additionally, an ultrasonic borehole probe is developed and constructed using phased arrays to further enhance the detection of potential cracks. Modelling and preliminary results from laboratory specimens prove the feasibility and potential of the directional response even in heterogeneous material such as concrete. Final investigations under in-situ conditions at the test site of the sealing structure are planned.
Overall, the project SealWasteSafe improves the construction material, multi-sensory monitoring, and ultrasound for quality assurance to allow for the development of safe nuclear sealing structures. Although the techniques are tailored for sealing structures within salt as a host rock, they are transferrable to a wider field of applications and alternative disposal conditions.
During the webinar "Dismantling and waste management: a playground for innovation" the innovative aspects of the PREDIS work package "innovations in cemented waste handling and pre-disposal storage" are presented. The webinar is jointly organised by Women in Nuclear Decommissioning (WINDI) and Women in Nuclear Innovation (WINI). The shown innovations include quality assurance (mainly NDE) and monitoring technologies with associated sensor systems, digital twin technology, and appropriate data handling including a decision framework. An overview of the on-going research in the field of pre-disposal waste management in the project Predis (WP7) is summarised.
Es werden Methoden zur sicherheitstechnischen Bewertung des mechanischen Verhaltens von Brennstabhüllrohren entwickelt, um sprödes Versagen in Belastungsszenarien nach längerfristiger trockener Zwischenlagerung zu verhindern. Unbestrahlte Brennstabhüllrohrsegmente aus der Zirkoniumlegierung ZIRLO® wurden hydriert und thermomechanischen Behandlungen unterzogen, um Hüllrohrmaterial durch Ausscheidung radialer Hydride zu verspröden. Die Proben mit radialen Hydriden wurden im Ring Compression Test (RCT) bei Raumtemperatur untersucht. Bereits bei kleinen Probenverformungen wurden abrupte Kraftabfälle mit instabiler Rissausbreitung durch das Netzwerk radialer Hydride beobachtet. Die Risslängen erreichten in einigen Fällen bis zu 90 % der Wanddicke. Der mikromechanische Versagensmechanismus ist Quasi-Spaltbruch in den Hydriden und Porenbildung, -wachstum und -vereinigung in der Zirkoniummatrix an den duktilen Brücken zwischen benachbarten Hydriden. Es wurde ein Finite-Elemente-Modell des Ring Compression Tests erstellt, um das Versagensverhalten zirkoniumbasierter Legierungen mit radialen Hydriden im RCT zu analysieren. Die Fließkurve für das elastisch-plastische Materialmodell des Matrixmaterials wurde durch eine inverse Finite-Elemente-Analyse ermittelt. Kohäsivzonenmodelle wurden benutzt, um die plötzlichen Kraftabfälle zu reproduzieren. Es wurde ein mikromechanisches Modell mit statistisch generierten Matrix-Hydrid-Verteilungen entwickelt, das zwischen sprödem Versagen entlang der Hydride und duktilem Versagen der Zirkoniummatrix in der Kohäsivzone unterscheidet. Für das jeweilige Materialverhalten wird ein lokales zweiparametriges Versagenskriterium auf Grundlage der Kohäsionsfestigkeit und der Separationsenergie bei einer festgelegten Form des Kohäsivgesetzes vorgeschlagen. Geeignete Kohäsivparameter lassen sich mit einer inversen Finite-Elemente-Analyse des Versagensverhaltens von hydrierten Proben im Ring Compression Test bestimmen. In den Simulationen lag der Schwerpunkt auf dem ersten Kraftabfall. Die Berechnungsergebnisse stimmen gut mit den RCT-Ergebnissen überein. Es konnte gezeigt werden, dass die Rissinitiierung und -ausbreitung stark von der speziellen Anordnung der radialen Hydride und Matrixbrücken in der Bruchzone abhängt. Die numerische Modellbildung wurde an Versuchen mit bestrahlten Proben aus der Zirkoniumlegierung M5® validiert.
The structural integrity of sealing structures in underground repositories for radioactive waste is a major concern regarding safe disposal. Non-destructive testing plays a very important role in the safety of this sealing structures. Ultrasonic pulse echo technology has been established in the construction industry as a non-destructive testing method for locating potential defects and material characterization of concrete structures. A borehole probe is an ultrasonic pulse-echo device used to investigate the internal structures of salt concrete. A prototype borehole probe is successfully demonstrated with the existing research borehole in the test site using a commercially available flaw detector. A novel phased array borehole probe that allows the beam to be focused based on the delay law was constructed and tested. The directivity measurements on the semi-cylindrical concrete sample agree well with the simulation results for the different focusing angles. In combination with geophysical reconstruction methods, the optimized radiation pattern of the probe is expected to improve the signal quality and thus increase the reliability of the imaging results. In addition, a unique large aperture ultrasound system (LAUS) is expected to provide enhanced images of possible cracks and embedded objects in in-situ structures. LAUS will be optimized to reliably detect objects at depths greater than 9 m and, at the same time, provide good resolution. Modeling studies show the potential to increase the distance between the transducer arrays further, and the same configuration has been successfully tested on a thick concrete wall. Overall, the paper shows that quality assurance using ultrasonic testing is crucial for constructing safe sealing structures to dispose of radioactive or toxic waste.
A new type of ultrasonic borehole probe is currently under development for the quality assurance of sealing structures in radioactive waste repositories using existing research boreholes. The goal is to examine the sealing structures made of salt concrete for possible cracks, delamination, and embedded objects. Earlier prototype probes use 12 or 16 individual dry point contact (DPC) horizontal shear wave transducers grouped into a transmitter and a receiver array, each made up of six or eight individual transducers. They are operated with a commercially available portable ultrasonic flaw detector used in the civil engineering industry. To increase the generated sound pressure of the borehole probe, the number of transducers in the novel probe is increased to 32. In addition, timed excitation of each probe is used to direct a focused sound beam to a specific angle and distance based on calculated time delays. Hence, the sensitive test volume is limited, and the signal-to-noise ratio of the received signals is improved. This paper presents the validation of the newly developed phased array borehole probe by beam computation in CIVA software and experimental investigations on a semi-cylindrical test specimen to investigate the directional characteristics. In combination with geophysical reconstruction techniques, an optimised radiation pattern of the probe is expected to improve the signal quality and thus increase the reliability of the imaging results.
This is of great importance for the construction of safe sealing structures needed for the disposal of radioactive or toxic waste.
Before the final disposal of radioactive wastes, various processes can be implemented to optimise the waste form. This can include different chemical and physical treatments, such as thermal treatment for waste reduction, waste conditioning for homogenisation and waste immobilisation for stabilisation prior to packaging and interim storage. Ensuring the durability and safety of the waste matrices and packages through performance and condition assessment is important for waste owners, waste management organisations, regulators and wider stakeholder communities. Technical achievements and lessons learned from the THERAMIN and PREDIS projects focused on low- and intermediate-level waste handling is shared here. The recently completed project on Thermal Treatment for Radioactive Waste Minimization and Hazard Reduction (THERAMIN) made advances in demonstrating the feasibility of different thermal treatment techniques to reduce volume and immobilise different streams of radioactive waste (LILW) prior to disposal. The Pre-Disposal Management of Radioactive Waste (PREDIS) project addresses innovations in the treatment of metallic materials, liquid organic waste and solid organic waste, which can result from nuclear power plant operation, decommissioning and other industrial processes. The project also addresses digitalisation solutions for improved safety and efficiency in handling and assessing cemented-waste packages in extended interim surface storage.
Validation of novel ultrasonic phased array borehole probe by using simulation and measurement
(2022)
Low-frequency ultrasonic testing is a well-established non-destructive testing (NDT) method in civil engineering for material characterization and the localization of cracks, reinforcing bars and delamination. A novel ultrasonic borehole probe is developed for in situ quality assurance of sealing structures in radioactive waste repositories using existing research boreholes. The aim is to examine the sealing structures made of salt concrete for any possible cracks and delamination and to localize built-in components. A prototype has been developed using 12 individual horizontal dry point contact (DPC) shear wave transducers separated by equidistant transmitter/receiver arrays. The probe is equipped with a commercially available portable ultrasonic flaw detector used in the NDT civil engineering industry. To increase the sound pressure generated, the number of transducers in the novel probe is increased to 32 transducers. In addition, the timed excitation of each transducer directs a focused beam of sound to a specific angle and distance based on the previously calculated delay time. This narrows the sensitivity of test volume and improves the signal-to-noise ratio of the received signals. In this paper, the newly designed phased array borehole probe is validated by beam computation in the CIVA software and experimental investigations on a half-cylindrical test specimen to investigate the directional characteristics. In combination with geophysical reconstruction methods, it is expected that an optimised radiation pattern of the probe will improve the signal quality and thus increase the reliability of the imaging results. This is an important consideration for the construction of safe sealing structures for the safe disposal of radioactive or toxic waste.
Langzeitversuche über 10 Jahre an Federkern-Metalldichtungen mit Aluminium- bzw. Silberummantelungen
(2023)
Federunterstützte Metalldichtungen mit Aluminium(Al)- oder Silber(Ag)-Ummantelung werden u.a. in Behältern für Wärme entwickelnde radioaktive Abfälle eingesetzt, da diese Dichtungen eine sehr gute Dichtheit gewährleisten sowie Langlebigkeit und Beständigkeit gegenüber erhöhten Temperaturen und radioaktiver Strahlung aufweisen. Auch wenn die Sicherheit solcher Behälter und der verwendeten Dichtungen vielfach belegt wurde, sind sie Gegenstand andauernder Forschung, etwa im Hinblick auf eine absehbar benötigte verlängerte Zwischenlagerdauer. Aus diesem Grund werden an der Bundesanstalt für Materialforschung und -prüfung (BAM) im Fachbereich 3.4 „Sicherheit von Lagerbehältern“ seit über 20 Jahren Versuche an solchen Metalldichtungen durchgeführt. Dabei sollen zusätzliche Erkenntnisse hinsichtlich der Sicherheitsreserven der Dichtungen in unterstellten Störfallszenarien (axiale Bewegung des Deckelsystems bzw. Aufweitung der Nutgeometrie) und insbesondere detailliertere Erkenntnisse zum Langzeitverhalten gewonnen werden.
The thermal expansion of polymeric parts can be an issue in many applications where the available space is limited, or exact dimensions of the part are required. For this study, a device was designed and built that allowed measuring the thermal expansion simultaneously in all three spatial directions on cubic samples with real-scale dimensions (78 mm edge length). The results are shown between 25 °C and 125 °C for two PE materials, one HMWPE and one tempered UHMWPE, for non-irradiated samples as well as cubes that have been irradiated with 100 and 400 kGy. The results measured with the new device were very similar to those measured with conventional thermo-mechanical analysis equipment and to literature data of UHMWPE. The HMWPE material shows a much larger thermal expansion coefficient in one direction compared to the other two directions during the first heating due to frozen stresses from the pressing step during material manufacturing. These stresses are mostly released by the expansion during the first heating, so that the expansion during the second heating is more uniform. The overall volumetric expansion is the same for both heating runs. By contrast, the tempered UHMWPE material shows no significant difference between first and second heating run, as the stresses from processing could already relax in the tempering step. The irradiation treatment does not affect the values significantly for the given test set-up.
At BAM, which is a federal institute for materials research and testing in Germany, it is one of our tasks to evaluate the safety of casks designed for transport and/or storage of radioactive material. This includes assessment of elastomeric seals applied in the casks. Besides examining the low-temperature behaviour and irradiation effects of elastomeric seals, it is our goal to estimate the service lifetime of the seals with regard to the requirements for long-term safety (40 years and more) of the containers. Therefore, we started an accelerated ageing programme with selected rubbers often used for seals (HNBR, EPDM and FKM) which are aged at four different temperatures (75 °C, 100 °C, 125 °C and 150 °C) up to five years. In order to assess sealability, O-rings were aged in compression by 25 % (corresponding to the compression during service) between plates as well as in flanges that allow leakage rate measurements. For comparison, uncompressed O-rings were aged as well. Further methods characterising seal performance are compression stress relaxation (CSR) reflecting the loss of sealing force of a compressed seal over time, and compression set (CS) which represents the recovery behaviour of a seal after release from compression. Additionally, material properties such as hardness, elastic modulus, glass transition temperature and viscoelastic loss factor as well as relaxation and recovery behaviour are examined in order to understand the underlying ageing mechanisms in each material. For obtaining results closely related to practical conditions, O-rings with a full-scale cord diameter of 10 mm were aged. However, this set-up can lead to heterogeneous aging caused by diffusion-limited oxidation (DLO) effects, resulting in distorted bulk properties such as compression stress relaxation and compression set. However, if DLO-affected data is excluded, extrapolations of CS data are possible using time-temperature shifts and Arrhenius graphs.
For selecting an appropriate end-of-lifetime criterion, leakage rate measurements were performed, since leakage rate is the only characteristic directly correlated to the performance of the sealing system. A significant increase in leakage rate was considered as the end of the lifetime. However, the O-rings remained leak tight under static conditions and even exhibited an improved, i.e. decreased leakage rate while other properties already indicated strong deterioration. These tests were extended by seal tests with fast partial decompression.
Overall an update will be given on the current investigations and results and the planned activities.
Übersicht über die gefahrgutrechtlichen Anforderungen zur Bauartprüfung und dem Qualitätsmanagement für Verpackungen für nicht zulassungspflichtige Versandstücke.
The friction at the interfaces between pieces of wood and steel is of crucial importance for the impact and energy absorption in impact limiters of transport casks for radioactive material. Here, the friction coefficient for the combinations wood-wood and wood-steel was measured in the temperature range between -40°C and 90°C. Results show decreasing friction with increasing temperature, ranging from 0.43 at -40°C to 0.22 for 90°C for wood-steel combinations and from 0.3 at -40°C to 0.24 at 90°C for a wood-wood combination.
This presentation summarizes the opbjectives and the progress of the EURATOM project PREDIS (pre-disposal management of radioactive waste) and its workpackage 7 in particular. Focus is on the development of wireless sensors for monitoring radioactive waste packages, measuring radiation, temperature, pressure and humidity.
Engineered barriers are a key element to enable safe nuclear waste disposal. One method currently under research for their construction is magnesia concrete applied in a shotcrete procedure. In this study, the ultrasonic echo method is evaluated as a means for quality assurance. Imaging of internal structures (backwall, boreholes) and defects, such as delamination, has successfully been achieved in the shotcrete. Additionally, detailed information about the potential cause of selected reflectors are obtained by phase analysis. In several test blocks of various sizes, no consistent concrete section boundaries have been found by ultrasonic imaging, which was verified by subsequent drilling and complementary tests. An experiment with artificial defects imitating cracks, air-filled voids, and material with lower density has been challenging and shows the limitations of the current methods. Although significant defects, such as a large delamination, are reliably identified, several smaller defects are not identified. Generally, ultrasonic imaging provides a suitable base as a mean for quality assurance during and after the construction of sealing structures. However, further developments are required to enhance the reliability of the method and a full validation is still pending. Still, the method has potential to increase the safety of nuclear waste repositories.
Für die Beförderung von umschlossenen radioaktiven Stoffen (Strahler) können Transporterleichterungen gelten, wenn das Strahlerdesign nachweislich unfallsicher ausgelegt ist und eine Zulassung als radioaktiver Stoff in besonderer Form vorliegt. Die in Deutschland zuständige Behörde für die Prüfung und Zulassung radioaktiver Stoffe in besonderer Form ist die Bundesanstalt für Materialforschung und –prüfung (BAM).
Eine oft langfristige Nutzung der Strahler kann eine alterungsbedingte Schwächung der Umschließung des radioaktiven Stoffes zur Folge haben. Jeder Strahler einer zugelassenen Bauart muss jedoch im Beförderungsfall zu jeder Zeit, auch nach längerer Nutzung, den vorgeschriebenen Prüfanforderungen genügen. Die Alterungsbewertung auf der Grundlage einer vom Antragsteller zu spezifizierenden Nutzungsdauer des radioaktiven Stoffes in besonderer Form ist seit langer Zeit Bestandteil des Zulassungsverfahrens in Deutschland. Ein von Deutschland eingebrachter Vorschlag im Rahmen der anstehenden Revision der IAEA Empfehlungen für die sichere Beförderung von radioaktiven Stoffen (SSR-6) soll die Lesbarkeit des Regelwerkes bezüglich der Alterungsbewertung verbessern, Klarheit über die damit verbundenen Anforderungen erzeugen und damit zu einer Harmonisierung der Zulassungsverfahren auf internationaler Ebene beitragen. In diesem Beitrag sollen wichtige Einflussfaktoren auf das Alterungsverhalten von radioaktiven Stoffen in besonderer Form aufgezeigt und die Notwendigkeit für die Spezifikation einer Nutzungsdauer als Grundlage für die Alterungsbewertung begründet werden. Der deutsche Vorschlag für die Revision des IAEA Regelwerks zum sicheren Transport radioaktiver Stoffe (SSR-6) wird vorgestellt und erläutert.
With our experiments we want to simulate a load situation which can result from a cask vertical drop test. The lid can lift a little for a very short moment and that can perhaps result in a little movement of the seal, a little rotation or a little lateral movement, so that the contact section can change. The aim of the project is to learn more about the sealing efficiency after the seal is compressed again. What we know from former component tests with non-aged seals is, that the leakage rate measured after one or more decompression /compression cycles with seal movement becomes significant higher. And this result is considered for specification of design leakage rates for release calculation in safety cases.
In the recent years we all have learned more about the significant influence of temperature and time on seal behavior and so the plan was to do similar experiments with aged seals, just to get an idea about seal behavior and achievable leakage rates.
What we did in detail is:
We compressed Helicoflex-seals of both design types, with Aluminum and Silver outer jackets, in testflanges and aged them in an oven at a temperature of 125°C for 3 months to produce seal properties comparable with properties after several years use at more typical operating temperatures between 90 and 100°C.
After this ageing procedure we opened the flanges completely, moved the seal a little to vary the contact area and compressed the flanges again. What we measured during all compression and decompression cycles was the load, the deformation and -as long as possible- the leakage rates.
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 the Second Research Coordination Meeting on Spent Fuel Characterization, BAM as leader of Task 3 reported on the status of the research work.
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 behaviour of spent nuclear fuel (SNF), irradiated cladding, fuel/cladding chemical interaction (FCCI) and ageing under conditions of extended interim storage, transportation and emplacement 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.
Untersuchung einer Schweißnaht mit verschiedenen Codes. Vergleiche von experimentellen und numerischen Ergebnissen
In order to demonstrate the safety of both spent fuel and the storage system, a good understanding of the processes that might cause deterioration is required. The International Atomic Energy Agency (IAEA) continued the coordinated research on Spent Fuel Performance Assessment and Research (SPAR) into a fourth phase starting in 2015 to evaluate fuel and materials performance under wet and dry storage and to assess the impact of interim storage on associated spent fuel management activities (such as handling and transport). BAM’s understanding and experiences in the field of long-term storage of spent fuel and especially spent fuel characterization are presented to assist the preparation of the final project report.
Embrittlement of spent fuel claddings during long-term dry interim storage - Current approach at BAM
(2019)
The integrity of fuel rods and cladding tubes must be demonstrated throughout the dry interim storage of spent nuclear fuel in Germany. For at least a relevant number of casks, an extension of interim storage between 40 and 100 years is foreseeable. The drying procedure for the fuel rods in conjunction with the thermomechanical storage conditions may change the microstructure of the cladding tubes and lead to their possible embrittlement. The failure probability might increase under mechanical load by handling procedures (transport after storage) on potentially embrittled cladding tubes of fuel rods especially at low temperatures after long-term storage. Results of ring compression tests on unirradiated pre-hydrided as well as irradiated samples with radial hydrides are investigated by finite element analyses to be able to describe the failure process and to identify failure criteria.
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.
The characterization of spent fuel and radioactive waste is essential in the evaluation of the ability of these materials to meet acceptance criteria. Prior to determining a method for spent fuel characterization, it is necessary to identify the type of data required to support safety and other analyses for the storage, transportation and final disposal of spent fuel. BAM’s understanding and capabilities in the field of spent fuel characterization are presented for the scope of a coordinated research project on the topic.
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.
Underground sealing structures are necessary to seal radioactive or toxic waste in underground repositories. A special version of MgO concrete is used in the mine in Teutschenthal and applied at a large-scale test with the aim to provide a low permeability barrier. The test structures (up to 10 m long) were created by shotcreting. Besides destructive tests, non-destructive ultrasonic measurements have been evaluated for quality assurance to monitor potential anomalies such as cracks, concrete inhomogeneities, and delamination. We show results of ultrasonic measurements with a multi-transducer system used at the front and from the side of the concrete structure. Images are obtained by synthetic aperture focusing techniques. The boundaries between concreting sections are not visible in the ultrasonic images systematically so that a successful concreting is assumed, which is confirmed by the low permeabilities observed. Several reflectors up to a depth of 1.2 m are identified and interpreted as potential damages in the concrete. Phase analysis of prominent reflectors reveal a negative impedance contrast thus indicating air filled voids, cracks or delamination. Boreholes through some of the identified reflectors are used to verify the results from ultrasonic measurements. Additionally, an experiment with incorporated artificial defects was undertaken to analyse the reliability of the applied methods. Overall, the conducted tests show the potential of ultrasonic measurements to detect critical anomalies. Despite challenges at small-scale structures (cm-order), large-scale anomalies can be identified. Consequently, implementing ultrasonic monitoring during and after the construction of concrete sealing structures has shown its potential as a tool for quality assurance, but needs further development and validation.