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- Radioaktive Stoffe (6)
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- Impact limiter (3)
- Safety assessment (3)
- Spent fuel (3)
- Cylindrical cask (2)
- Explicit dynamics (2)
- Gefahrgut (2)
- IAEA puncture drop test (2)
Organisationseinheit der BAM
CASTOR-Behälter sollen den Auswirkungen schwerster Unfälle standhalten und müssen dabei ihre Sicherheitsfunktionen beibehalten. Dazu zählen der sichere Einschluss des radioaktiven Inventars mit nachgewiesener Behälterintegrität und -dichtheit, die ausreichende Abschirmung der radioaktiven Strahlung, die Ableitung der von den Brennelementen ausgehenden Wärme und der Ausschluss des Entstehens einer nuklearen Kettenreaktion, d.h. die Kritikalitätssicherheit sind zu gewährleisten.
Zum Nachweis dieser Fähigkeiten werden die Behälter gegen Stoß, Aufprall, Durchstoßen, Feuer sowie beim Eintauchen in Wasser geprüft. Durch diese Tests werden Beanspruchungen aus potentiellen schweren Unfällen abdeckend simuliert. Die hierfür erforderlichen experimentellen Tests, aber auch die Prüfung und Entwicklung komplexer rechnerischer Simulationen des Behälterverhaltens unter Anwendung der Finite-Elemente-Methode erfolgen bei der Bundesanstalt für Materialforschung und -prüfung (BAM).
Die Basis bildet das Regelwerk der Internationalen Atom- und Energieorganisation (IAEO) zum sicheren Transport von radioaktiven Stoffen, welches international harmonisierte Schutzziele und Prüfanforderungen definiert. Die seit 60 Jahren praktizierte stetige Weiterentwicklung der Sicherheitsanforderungen trägt aktuellen Erkenntnissen, Erfahrungen und Risikoanalysen Rechnung. Die Brennelementbehälter werden mechanischen Prüfungen sowie Brand- und Wasserdruckprüfungen unterzogen. Hierdurch wird nachgewiesen, dass sie den Auswirkungen eines schweren Unfalls standhalten können.
Nur Behälter die zweifelsfrei die kumulativen mechanischen und thermischen Tests und auch die Wasserdruckprüfung mit Erfüllen aller Schutzziele bestehen, bekommen eine Zulassung, die es erlaubt die Behälter zu transportieren.
When storage of spent nuclear fuel or high level waste is carried out in dual purpose casks (DPC), the effects of aging on safety relevant DPC functions and properties have to be managed in a way that a safe transport after the storage period of several decades is capable and can be justified and certified permanently throughout that period. The effects of aging mechanisms (e.g. radiation, different corrosion mechanisms, stress relaxation, creep, structural changes and degradation) on the transport package design safety assessment features have to be evaluated. Consideration of these issues in the DPC transport safety case will be addressed. Special attention is given to all cask components that cannot be directly inspected or changed without opening the cask cavity, like the inner parts of the closure system and the cask internals, like baskets or spent fuel assemblies. The design criteria of that transport safety case have to consider the operational impacts during storage. Aging is not the subject of technical aspects only but also of intellectual aspects, like changing standards, scientific/technical knowledge development and personal as well as institutional alterations. Those aspects are to be considered in the management system of license holders and in appropriate design approval update processes. The paper addresses issues that are subject of an actual International Atomic Energy Agency TECDOC draft 'Preparation of a safety case for a dual purpose cask containing spent nuclear fuel'.
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. In this paper, the mechanical behavior of high burn-up spent fuel assemblies (> approx. 50 GWd/tHM, value averaged over the fuel assembly) under transport conditions is 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. Additionally, 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. Considerations and knowledge gaps for the transport after extended interim storage are issues of growing interest. In this context, practical approaches are discussed based on the experience of BAM within the safety assessment of packages approved for transport of spent nuclear fuel.
Dicht und sicher abgeschirmt
(2014)
Transport casks for radioactive materials have to withstand the 9 m drop test, 1 m puncture drop test and dynamic crush test with regard to the mechanical requirements according to the IAEA regulations. The safety assessment of the package can be carried out on the basis of experimental investigations with prototypes or models of appropriate scale, calculations, by reference to previous satisfactory safety demonstrations of a sufficiently similar nature or a combination of these methods. Computational methods are increasingly used for the assessment of mechanical test scenarios. However, it must be guaranteed that the calculation methods provide reliable results. Important quality assurance measures at the Federal Institute for Materials Research and Testing are given concerning the preparation, run and evaluation of a numerical analysis with reference to the appropriate guidelines. Hence, a successful application of the finite element (FE) method requires a suitable mesh. An analysis of the 1 m puncture drop test using successively refined FE meshes was performed to find an acceptable mesh size and to study the mesh convergence using explicit dynamic FE codes. The FE model of the cask structure and the puncture bar is described. At the beginning a coarse mesh was created. Then this mesh was refined in two steps. In each step the size of the elements was bisected. The deformation of the mesh and the stresses were evaluated dependent on the mesh size. Finally, the results were extrapolated to an infinite fine mesh or the continuous body, respectively. The uncertainty of the numerical solution due to the discretisation of the continuous problem is given. A safety factor is discussed to account for the uncertainty.
In the approval procedure of transport packages for radioactive materials, the competent authority mechanical and thermal safety assessment is carried out in Germany by BAM Federal Institute for Materials Research and Testing. The combination of experimental investigations and numerical calculations in conjunction with materials and components testing is the basis of the safety assessment concept of the BAM. Among other mechanical test scenarios, a 1 metre drop test onto a steel bar has to be considered for the application of the hypothetical accident conditions to Type B packages according to IAEA regulations. Within the approval procedure for the new German package design of the HLW cask CASTOR® HAW 28M, designed by GNS Gesellschaft für Nuklear-Service Germany, a puncture drop test was performed with a half-scale model of the cask at -40°C. For independent assessment and to control the safety analysis presented by the applicant, BAM developed a complex finite element (FE) model for a dynamical ABAQUS/ExplicitTM analysis. This paper describes in detail the use of the FE method for modelling the puncture drop test within an actual assessment strategy. At first, investigations of the behaviour of the steel bar were carried out. Different friction coefficients and the material law of the bar were analysed by using a 'rigid-body' approximation for the cask body. In the next step, a more detailed FE model with a more realistic material definition for the cask body was developed. The validation of calculated strains was carried out by comparison with the results of the strain gauges located at the relevant points of the cask model. The influence of the FE meshing is described. Finally, the validated FE half-scale model was expanded to full-scale dimension. Scaling effects were analysed. The model was used for safety assessment of the package to be approved.
German approach and feedback on experience of transportability of SNF packages after interim storage
(2014)
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 package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the recently revised 'Guidelines for dry interim storage of irradiated fuel assemblies and heat-generating radioactive waste in casks' by the German Waste management Commission. For transport on public routes between or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfil the specifications of the transport approval or other sufficient properties, which satisfy the proofs for the compliance of the safety objectives at that time. In recent years, the validation period of transport approval certificates for manufactured, loaded and stored packages were discussed among authorities and applicants. A case dependent system of 3, 5 and 10 years was established. There are consequences for the safety cases in the Package Design Safety Report, including evaluation of long term behaviour of components and specific operating procedures of the package. The present research and knowledge concerning the long term behaviour of transport and storage cask components have to be consulted as well as experiences from interim cask storage operations. Challenges in the safety assessment are e.g. the behaviour of aged metal and elastomeric gaskets under IAEA test conditions to ensure that the results of drop tests can be transferred to the compliance of the safety objectives at the time of transport after the interim storage period. Assessment methods for the material compatibility, the behaviour of fuel assemblies and the aging behaviour of shielding parts are issues as well. This paper describes the state of the art technology in Germany, explains recent experience on transport preparation after interim storage and points out arising prospective challenges.