Kerntechnische Entsorgung
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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.
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).