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Organisationseinheit der BAM
Transport packages for spent fuel have to meet the International Atomic Energy Agency requirements for different transport conditions. Physical state of spent fuel and fuel rod cladding as well as geometric configuration of fuel assemblies are important inputs for the evaluation of package capabilities under these conditions. In this paper, the mechanical behaviour of high burn-up spent fuel assemblies under transport conditions is analysed 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 (fine cracks or complete breakage), which can cause release of gas, volatiles, fuel particles or fragments have to be properly considered in these assumptions. In view of the complexity of interactions between the fuel rods as well as between 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 pellets, especially for the high burn-up fuel, make the analysis more complicated additionally. In this context some practical approaches based on experiences of BAM Federal Institute for Material Research and Testing within safety assessment of packages for transport of spent fuel are discussed.
Investigation of the internal impact during a 9 m drop test of an accident-safe waste package
(2021)
The safety assessment of packages for the transport of radioactive material follows the IAEA regulations and guidance. The specified regulatory tests cover severe accidents and demonstrate the package containment system integrity.
Special attention must be drawn to the behaviour of the content which could move inside the package due to unpreventable gaps caused by the loading procedure and the structure of the content. A possible internal impact of the content which occurs during the drop tests onto the lid system is investigated. The IAEA regulations SSR-6 and the Guidance SSG-26, revised recently, consider input from Germany and France related to the significance of internal gaps.
In the context of a waste package design assessment, a model was equipped with a representative content to conduct a drop test with an internal impact. The weight and kinetic impact of this content covered all possible real contents. The objective of the test was to maximize the load onto the lid system and to prove the mechanical integrity by complying with the required leak tightness. The test was conducted
conservatively at a package temperature lower than -40 °C at the BAM Test Site Technical Safety.
This paper gives an overview of efforts to address internal gaps and their consequences, and the BAM efforts with the implementation of this topic into IAEA regulations and guidance material. The paper then focuses on the conduction of a drop test and investigation of internal component impact.
The requirements of the IAEA safety standards for Type B(U) packages include the thermal test as part of test sequences that represents accident conditions of transport. In comparison to mechanical tests, e.g., 9 m drop onto an unyielding target with short impact durations in a range of approximately 10 ms to 30 ms, the extended period of 30 min is defined in regulations for exposure of a package to a fire environment. Obviously, the required containment capability of the package has to be ensured not only after completing the test sequence but also over the course of the fire test scenario.
Especially, deformations in the sealing area induced by the non-uniform thermal dilation of the package can affect the capability of the containment system. Consequently, thermo-mechanical analyses are required for the assessment.
In this paper some aspects of finite element analysis (FEA) of transport packages with bolted closure systems under thermal loading are discussed. A generic FE model of a cask is applied to investigate the stress histories in the bolts, lid, and cask body as well as the deformations in the sealing area and the compression conditions of the gasket. Based on the parameter variations carried out, some recommendations in regard to modeling technique and results interpretation for such kind of analyses are finally given.
The regulatory compliance of the containment system is of essential importance for the design assessment of transport packages for radioactive materials. The requirements of the IAEA transport regulations SSR-6 for accident conditions implies high load on the containment system of Type B(U) packages. The integrity of the containment system has to be ensured under the mechanical and thermal tests.
The containment system of German transport packages for spent nuclear fuel (SNF) and high level waste (HLW) usually includes bolted lids with metal gaskets. BAM Federal Institute for Materials Research and Testing as the German competent authority for the mechanical and thermal design assessment of approved transport packages has developed the guideline BAMGGR 012 for the analysis of bolted lid and trunnion systems.
According to this guideline the finite element (FE) method is recommended for the calculations. FE analyses provide more accurate and detailed information about loading and deformation of such kind of structures. The results allow the strength assessment of the lid and bolts as well as the evaluation of relative displacements between the lid and the cask body in the area of the gasket groove.
This paper discusses aspects concerning FE simulation of lid systems for SNF and HLW transport packages. The work is based on the experiences of BAM within safety assessment procedures.
The issues considered are the assessment methods used in the BAM-GGR 012 for bolted lid systems along with the nominal stress concept which is applied for bolts according to that guideline. Additionally, modeling strategies, analysis techniques and the interpretation of the results are illustrated by the example of a generalized bolted lid systems under selected accident conditions of transport.
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.
The containment systems of transport casks for spent fuel and high radioactive waste usually include bolted lids with metallic or elastomeric seals.
The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport can have a significant effect on the leak tightness of such containment system.
Scaled cask models are often used for providing the required mechanical and thermal tests series. Leak tests have been conducted on those models.
It is also common practice to use scaled component tests to investigate the influence of deformations or displacements of the lids and the seals on the standard leakage rate as well as to study the temperature and time depending alteration of the seals.
In this paper questions of the transferability of scaled test results to the full size design of the containment system will be discussed.
Numerical approach to determine the correct puncture bar length for the IAEA puncture bar drop test
(2018)
Among other mechanical tests the 1 meter drop onto a steel puncture bar shall be considered for accident safe packages for the transport of radioactive material. According with the IAEA regulations “the bar shall be of solid mild steel of circular section, 15.0 ± 0.5 cm in diameter and 20 cm long, unless a longer bar would cause greater damage…”. The most damaging puncture bar length can be estimated by iterative processes in numerical simulations. On the one hand, a sufficient puncture bar length has to guarantee that shock absorbers or other attachments do not prevent or reduce the local load application to the package, on the other hand, a longer and thus less stiff bar causes a smaller maximum contact force. The contrary influence of increasing puncture bar length and increasing effective drop height shall be taken into account if a shock absorber is directly placed in the target area. The paper presents a numerical approach to identify the bar length that causes maximum damage to the package. Using the example of two typical package masses the sensitivity of contact forces and puncture bar deformations to the initial length are calculated and assessed with regard to the international IAEA package safety requirements.
Transport and storage of radioactive materials are performed in countries with policy of either closed or open nuclear fuel cycle. The related technologies have been established by accumulation of experiences and researches including demonstrative tests using full scale or scale models and analyses. Those are essential before commercialization, but are often costly and time consuming. Such demonstrative works should not be repeated meaninglessly, but can be shared through this kind of book and used by readers and the future generations to advance the technology effectively. This book systematically provides findings from lots of valuable researches on safety of transport and storage of radioactive materials under normal and accident conditions that have an impact on basis of safe regulations, designs, and operations.