Angewandte Physik
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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.
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.
Es werden die Zusammenhänge des Druckvolumenprodukts und der Konsequenz bei Druckgasen insbesondere für Wasserstoff dargestellt. Entsprechend werden Stellungnahmen zu verschiedenen Normungsprojekten abgeleitet und perspektivische Entwicklungen mit dem resultierenden Bedarf in der Normung dargestellt. Außerdem wird am Beispiel ein Problem in der praktischen Beurteilung von Speicherpopulationen für Batteriefahrzeuge diskutiert.
Packages for the transport of spent nuclear fuel shall meet the International Atomic Energy Agency regulations to ensure safety under different Transport conditions. The physical state of spent fuel and the fuel rod cladding as well as the geometric configuration of fuel assemblies are important inputs for the evaluation of package capabilities under these conditions. Generally, the mechanical behavior of high burn-up spent fuel assemblies under Transport conditions shall be analyzed with regard to the assumptions which are used in the containment and criticality safety analysis. In view of the complexity of the interactions between the fuel rods as well as between the fuel assemblies, basket, and cask containment, the exact mechanical analysis of such phenomena is nearly impossible. The gaps in information concerning the material properties of cladding and pellet behavior, especially for the high burn-up fuel, make the analysis more complicated additionally. As a result, enveloping analytical approaches are usually used by BAM within the safety assessment of packages approved for transport of spent nuclear fuel.
To justify the safety margins of such approaches additional analyses are necessary. In this paper, numerical simulations of a segment of a spent fuel assembly are presented. The segment modeled represents the part of a generalized BWR fuel assembly between two spacers. Explicit dynamic finite element calculations are performed to simulate the spent fuel behavior under regulatory defined accident conditions of transport. A beam element formulation is used for the modeling of the fuel rods representing the compound consisting of claddings and fuel pellets. The load applied is gathered from experimental drop tests with spent fuel casks performed at BAM. A hot cell bending test performed at JRC Karlsruhe is the basis for obtaining the material behavior of the fuel rods. The material properties are determined by
simulating the test setup of JRC and optimizing the results to fit the
experimental load deflection curve. The simulations of the fuel Assembly segment are used to get a better understanding about the loads on fuel rods under accident conditions of transport.
To increase the competitiveness of jacket substructures compared to monopiles a changeover from an individual towards a serial jacket production based on automated manufactured tubular joints com-bined with standardized pipes has to be achieved. Therefore, this paper addresses fatigue tests of automatical-ly welded tubular X-joints focusing on the location of the technical fatigue crack. For this X-joint, the detect-ed location of the technical crack is then compared to numerical investigations predicting the most fatigue prone notch considering the structural stress approach as well as the notch stress approach. Additionally, the fatigue prone hot spot according to both approaches is compared for a typical offshore jacket double-K-joint to emphasize the significance of the presented outcomes for the existing offshore structures. Besides, the welding process of the automated manufactured tubular X-joints is presented.
To increase the competitiveness of jacket substructures compared to monopiles a changeover from an individual towards a serial jacket production based on automated manufactured tubular joints com-bined with standardized pipes has to be achieved. Therefore, this paper addresses fatigue tests of automatical-ly welded tubular X-joints focusing on the location of the technical fatigue crack. For this X-joint, the detect-ed location of the technical crack is then compared to numerical investigations predicting the most fatigue prone notch considering the structural stress approach as well as the notch stress approach. Additionally, the fatigue prone hot spot according to both approaches is compared for a typical offshore jacket double-K-joint to emphasize the significance of the presented outcomes for the existing offshore structures. Besides, the welding process of the automated manufactured tubular X-joints is presented.
To increase the competitiveness of jacket substructures compared to monopiles a changeover from an individual towards a serial jacket production based on automated manufactured tubular joints combined with standardized pipes has to be achieved. Therefore, this paper addresses fatigue tests of automatically welded tubular X-joints focusing on the location of the technical fatigue crack. The detected location of the technical crack is compared to numerical investigations predicting the most fatigue prone notch considering the structural stress approach as well as the notch stress approach. Besides, the welding process of the automated manufactured tubular X-joints is presented.
To increase the competitiveness of jacket substructures compared to monopiles a changeover from an individual towards a serial jacket production based on automated manufactured tubular joints combined with standardized pipes has to be achieved. Therefore, this paper ad-dresses the welding process of automated manufactured tubular X-joints. Additionally, fatigue tests of the automatically welded tubular joints are presented focusing on the location of the technical fatigue crack. The detected location of the technical crack is compared to numerical investigations predicting the most fatigue prone notch considering the structural stress approach as well as the notch stress approach.
Packages for the transport of spent nuclear fuel shall meet the International Atomic Energy Agency regulations to ensure safety under different transport conditions. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of fuel assemblies are important inputs for the evaluation of the package capabilities under these conditions. The mechanical behavior of spent fuel assemblies shall be analyzed with regard to the assumptions which are used in the containment and criticality safety analysis. The package as a mechanical system is characterized by a complex set of interactions, e. g. between the fuel rods within the assembly as well as between the fuel assemblies, the basket, and the cask containment. This complexity makes it difficult to assume appropriate mechanical loads for the spent fuel inside the package. Another challenge is the assumption of material properties which represent the variation of the fuel rods regarding cladding material, burn-up and the operation history sufficiently. The objective of this paper is to give an overview about the current approaches and research for the evaluation of spent fuel behavior within the package design approval procedure. In this context, analytical, numerical and experimental results are discussed. The presented work is achieved within a collaboration of BAM with national and international partners.