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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.
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.
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.