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- Spent fuel (3)
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- Aktivitätsfreisetzung (1)
- Arrhenius equation (1)
The German approach of storing spent nuclear fuel in dual purpose casks until final disposal is based on individually licensed storage sites concerning specific nuclear inventories and dual purpose cask types. The currently approved license period is maximum 40 years.
With nuclear policy changes including the phase-out decision of 2011 and a new approach for the HLW repository siting process set up 2013 by law also the spent fuel and HLW management strategy has to be adapted. This includes the timely development of disposal container concepts as part of the overall repository siting process and the need for longer interim storage periods until final disposal. Other aspects affect new regulations about the financing of the nuclear waste management by the industry and the public sector, the restructuring of responsible authorities in 2016/2017 and the adaption of R&D funding programs addressing also pre-disposal (or extended interim storage) issues.
BAM as a Federal Research Institute has been responsible for or in charge of container safety Evaluation for transport and storage casks for radioactive waste for several decades and started addressing potential Extended interim storage issues already several years ago. This included participation in international projects at the IAEA or in the U.S. Extended Storage Collaboration Program (ESCP) as well as long term laboratory tests with metal seals, elastomeric seals and polymers for neutron shielding purposes. Test results of BAM are published regularly and gained data provide valuable information about thermo-mechanical and radiological degradation effects as well as the long term Performance of respective cask components. Regarding the reliable extrapolation and prediction of the long term Performance of cask materials and components for time periods of up to 80 or even 100 years the continuation and extension of laboratory tests and the development of analytical and numerical simulation models and tools is essential and under way.
In Germany the concept of dry interim storage of spent fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years. The casks being used have an approved transport package design in accordance with the international IAEA transport regulations as well. The storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage.
Besides consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case.
Not only the technical objects are subject of aging but also the safety demonstration basis is subject of “aging” due to possible changes of regulations, standards and scientific/technical knowledge.
To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the Package Design Safety Report (PDSR), in connection with periodic renewals of package design approval certificates, have to be carried out.
German packages for the transport of spent nuclear fuel are assessed with respect to specific transport conditions which are defined in the safety regulations of the International Atomic Energy Agency.
In general, gastight fuel rods constitute the first barrier of the containment system. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package safety under transport conditions. The objective of this paper is to discuss the methodologies accepted by German authority BAM for the evaluation of spent fuel behavior within the package design approval procedure.
Specific test conditions will be analyzed with regard to assumptions to be used in the activity release and criticality safety analysis. In particular the different failure modes of the fuel rods, which can cause release of gas, volatiles, fuel particles or fragments, have to be properly considered in these assumptions.
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 together with the limited knowledge about the material properties and the variation of the fuel assemblies regarding cladding material, burn-up and the operation history makes an exact mechanical analysis of the fuel rods nearly impossible.
The application of sophisticated numerical models requires extensive experimental data for model verification, which are in general not available. The gaps in information concerning the material properties of cladding and pellets, especially for the high burn-up fuel, make the analysis more complicated additionally, and require a conservative approach.
In this context some practical approaches based on experiences by BAM within safety assessment of packages for transport of spent fuel will be discussed.
Ongoing research activities to investigate SNF mechanical behavior in view of gas and fissile material release under transport loads are presented.
Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the regulations of the International Atomic Energy Agency. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important Inputs for the evaluation of the package capabilities under These conditions. Cracks or failures in the fuel rod cladding can cause the release of gas, volatiles or fuel particles into the cavity. The amount of substances in the cavity has to be considered in the assessment of the activity release and criticality safety.
The mechanical analysis of the compound system formed by the fuel rod cladding and the spent fuel pellets is very difficult due to the limited knowledge of the material properties and the insufficient understanding of the interaction between pellets and cladding and between adjacent pellets. The variation of fuel assembly properties regarding cladding material, burn-up and the history of usage makes reliable predictions of the fuel rod behavior even harder.
For a better understanding about the behavior of spent fuel rods, JRC and BAM have started a joint research project. In this context, JRC has developed a test device which allows quasi-static 3-point-bending test on fuel rod segments in the hot cell. The loads are applied with respect to the boundary conditions of the activity release assessment. This paper deals with the numerical calculation of a single fuel rod segment under bending load. The aim is to identify the governing mechanical parameters by the variation of constitutive assumptions, contact conditions, inner constraints, etc. This knowledge helps for the interpretation of the experimental results. Furthermore, the improved understanding about the behavior of the cladding-pellets system will be beneficial for the assessment of spent fuel transport conditions.
Piezoresistive accelerometers use a strain-sensing element, generally made of semiconductor material, e.g., silicon to convert the mechanical motion into an electrical signal. This element is usually designed in form of a cantilever beam loaded with a mass. Acceleration causes bending of the beam, which produces a change of electrical resistance proportional to the applied acceleration.
Main advantages of piezoresistive accelerometers in comparison to other types, e.g., piezoelectric and capacitive, is their robust and highly dynamic behavior, which qualifies them for application in high impact shock applications. Mechanical damping is typically implemented with silicon oil in a way that the output signal is undistorted over a wide frequency range. These characteristics principally qualify them for the application in drop tests carried out at BAM, for which they are calibrated over the frequency range from 1 to 4 kHz. However, using silicon oil for damping, has the drawback of temperature dependent change of its viscosity, leading to temperature dependent deviation of the accelerometer’s sensitivity.
This study presents experimental results of the temperature behavior of a piezoresistive accelerometer with a dynamic range up to ±5000 g.
This type of accelerometer is applied for drop tests which are partially performed at temperatures of -40 or +100 °C.
Trunnion systems of packages for the transport of radioactive materials have to guarantee the safe handling of package during crane operations (lifting, tilting) and to secure package tie down to the transport vehicle, if the trunnions are used as attachment points during transport. The design of trunnions is based on IAEA Regulations SSR-6, the supporting Advisory Material SSG-26 and further appropriate technical standards and/or guidelines. For package approval procedures in Germany the guideline BAM-GGR 012 has to be applied. If a package is handled inside a german nuclear power plant, the standard KTA 3905 has to be fulfilled additionally.
In this paper the requirements of KTA 3905 concerning the trunnion systems as load attaching points (LAP) are discussed in connection with the recommendations in the guideline BAM-GGR 012. This guideline is prepared at BAM Federal Institute for Materials Research and Testing for analysis and assessment of bolted lid and trunnion systems of Type B(U) transport packages. The Quality assurance questions concerning trunnion systems are discussed as well.
For the application of the time-temperature superposition principle a suitable relation is needed to describe the time-temperature shift factor α. Therefore, the Arrhenius equation is widely used due to its simple form and often leads to suitable results. Where, the Arrhenius equation presents a linear relation for the temperature-dependent shift factor in logarithmic scale ln(α) with the absolute inverse temperature (1/ϑ). However, in cases with a large temperature range which eventually include more complex reaction processes, the functional relation between ln(α) and (1/ϑ) is nonlinear in the ‘Arrhenius plot’. In those cases, the monotone change of the nonlinear range in the ‘Arrhenius plot’ can be interpreted as a transient range between two approximately linear or constant regions. An extended application of the modified Arrhenius equation from Nakamura (1989) is presented in this study for this transient range. The introduced method was applied to describe the time-temperature equivalence in the relaxation analysis of restoring seal force of metal seals, which are used in lid-systems of transport and interim storage casks for radioactive materials. But, the method is widely valid and can be used for different objectives which are characterized by thermorheologically simple behavior with nonlinear sensitivity to inverse temperature.
Drop testing of a container for the storage. transport and disposal of intermediate level waste
(2017)
Impact tests were performed, as part of a corresponding container’s substantiation, during design development of a shielded container. The container will be used for storage, transport, and disposal of intermediate level waste in the UK.
The mechanical test program comprised a 9m free drop test onto an unyielding target in a container long lid edge down orientation at ambient conditions. Further, a 0,5m free drop test onto a punch target was performed. Here, the container was orientated with the lid downwards in a way that the punch, a mild steel bar, impacts a filter lid.
The test specimen was instrumented with strain gauges and accelerometers for the drop tests. Transient strains at selected points of the inner and outer container walls, at the shielding lid, as well as at the lid bolts were measured during the container’s impact. Furthermore, decelerations of the container body, container lid, and the skip were measured. The complex geometrical changes of the container due to impact were determined by optical 3d- deformation measures using the projected fringes method in combination with multi-image photogrammetry.
This paper summarizes the performance of the drop tests and various drop test results in context with the design development as well as aspects regarding the associated Finite Element (FE) analyses and post-test evaluation.
The German nuclear waste management strategy consists of dry interim storage of spent fuel inside dual purpose casks and its subsequent direct final disposal in a deep geological repository. Interim storage is limited to 40 years so far, but will have to be extended until repository site selection and operation is concluded, potentially up to 100 years. In this context, research on the long term performance of safety relevant cask components and fuel rod integrity during such extended periods of time is of outmost importance. The barrier function of the fuel cladding depends on its integrity under operational and accidental conditions along with storage and subsequent transportation and is mainly affected over time by altering of the cladding microstructure and by thermo-mechanical conditions during reactor operation and subsequent wet and dry storage.
This paper focuses on failure mechanisms regarding cladding embrittlement caused by physical phenomena concerning hydrogen dissolution and precipitation of hydrides in the zirconium matrix. Within this context, potential cladding integrity issues under extended interim storage conditions followed by transportation are emphasized.
Delayed hydride cracking (DHC) is a well-known time-dependent temperature-activated phenomenon failure mechanism of the fuel cladding during reactor operation. The phenomenon consists of hydrogen diffusion and hydride precipitation at the tip of an initial crack and the propagation of the crack to an approximate distance equal to the length of the hydride. This failure mechanism is temperature-activated phenomenon and enhances along with thermal cycling of the fuel cladding, e. g. during reactor shutdowns. DHC shows limited relevance under quasi static storage conditions with continuously lowering temperatures. However, the literature points that DHC may occur when the temperature during storage decreases to a certain level after some decades and in combination with stress concentrations in the cladding material.
When spent fuel is transferred to dry storage by vacuum drying, the cladding temperature raises up to peak temperatures of ~370°C. Under these conditions, hydrides dissolute into the zirconium matrix up to the solubility limit defined by the peak temperature and this amount of hydrogen can precipitate in a radial re-oriented fashion due to the temperature decrease during storage. Cladding materials with radially precipitated hydrides show significant embrittlement. Radially oriented hydrides are perpendicularly oriented to hoop stresses representing the most critical configuration. Brittle failure of the cladding then occurs by crack propagation through radially oriented hydrides and finally through the zirconium matrix. The susceptibility to radial hydride precipitation depends on cladding material type and microstructure, hydrogen content, pre-drying hydride distribution, irradiation conditions, and temperature and stress histories during drying and storage operations.
This paper addresses major considerations concerning spent fuel cladding embrittlement due to hydride precipitation with regard to the boundary conditions of dry spent fuel storage in Germany. Analytical, numerical, and experimental approaches are to be discussed by BAM in order to identify the specific needs for future R&D work in that area with the purpose to provide the necessary data base for proper safety demonstration and evaluation along with future extended storage licensing procedures.
With the clear expectation of the need for extending spent fuel storage in Germany beyond the initial license period of maximum 40 years additional material data and knowledge about the long term performance and integrity of casks and internals like fuel rods will be required to meet major safety goals and regulatory requirements.
Based on a broad knowledge base about fuel cladding behavior, degradation mechanisms and related potential failure mechanisms, e. g. induced by reoriented Hydride precipitation, as illustrated in this paper, a specific gap analyses for the German situation regarding fuel conditions and cladding materials should identify the specific knowledge gaps to be addressed and closed in the near future. This is of specific interest as all spent fuel assemblies are going to be transferred to dry cask storage with the next decade and as a consequence of the German nuclear phase-out decision.
Within several decades, BAM as a Federal materials research institute gathered a broad range of expertise in the safety evaluation of materials, components and casks under operational and accidental thermal and mechanical conditions. This paper elucidates selected Basic considerations regarding technical and scientific knowledge about the long term aging performance and potential failure mechanisms of fuel cladding materials.
Under consideration of the national R&D funding policy which has been adapted to the extended interim storage or pre-disposal needs and the competences of BAM, its future R&D strategy is explained and some R&D objectives have already been identified. BAM will continue and expand ist engagement in the area of investigating fuel cladding material behavior in the long term and developing validated simulation and analyses tools in collaboration with national and international partners. R&D outcomes shall improve the knowledge-base to close identified knowledge gaps in the area of demonstrating fuel cladding integrity and to provide crucial information with regard to future safety assessments for the extended interim spent fuel storage.