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Eingeladener Vortrag
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The design of a special form radioactive material has to resist a severe transport accident without undue loss or dispersal of radioactive material.
Safety assessment by authorities competent for design approval has to include besides the required test program (impact, percussion, bending and heat test) also the evaluation of the quality management system for design, manufacture, testing, documentation, use, maintenance and inspection.
These quality assurance measures have to assure that every specimen of the approved design is produced in the same verified quality and every specimen must be able to survive the severe mechanical and thermal tests without undue loss or dispersal of radioactive material at any time of its working life.
All important aspects in the design approval procedure by BAM as the competent authority for approvals of special form radioactive material in Germany are summarized in a guideline published in 2014. This paper will give additional explanations to some aspects in safety assessment, e.g.:
the applicability of leak test methods and the need to consider ageing aspects.
Es wird über die Weiterentwicklung des ASME (American Society of Mechanical Engineers) – Regelwerks berichtet. Relevant für das Design von Behältern ist ASME BPVC (ASME Boiler and Pressure Vessel Code) Section III Division 3. In den ASME BPVC in der Fassung von 2015 wurden dehnungsbasierte Bewertungskriterien aufgenommen. Eine dehnungsbasierte Bewertung erfordert die genaue Kenntnis der aufgetretenen Dehnungen, d.h. speziell möglichst genaue numerische Berechnungen der Behälterbeanspruchungsszenarien. Daher sind regulatorische Vorgaben notwendig für die Vorbereitung, Durchführung und Auswertung numerischer Simulationen. In einer Arbeitsgruppe (ASME Special Working Group on Computational Modeling for Explicit Dynamics) werden dazu Leitlinien für die Anwendung numerischer Berechnungsverfahren entwickelt. Im Vortrag werden diese Leitlinien vorgestellt.
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.
It is important to evaluate the effect of thermal ageing on the sealing performance of metal gaskets under high temperature for long-term usage. Therefore, in order to gain representative data for this kind of metal-sealed lid System, BAM is currently performing laboratory tests of different gasket types with aluminum and silver jackets at three different temperatures under static conditions up to four years so far, using test flanges for gaskets with full scale cross section diameter but much smaller outer diameter. On the other hand, in order to investigate the applicability of the numerical methodology to evaluate the long-term behavior of the metal gaskets, such as a correlation between seal pressure force and holding time, CRIEPI is developing a modeling method including material tests (tensile and creep tests) at high temperature. In this paper, the applicability of the finite element method (ABAQUS) to predict the recovery displacement and residual seal pressure force of the gasket complex was verified by comparing the calculated values with BAM’s laboratory test results under the joint research agreement between BAM and CRIEPI.
Packages for the transport of radioactive material are often equipped with impact limiters consisting of wood, encapsulated by steel sheets. These impact limiters shall ensure the transport cask meets the mechanical and thermal IAEA regulatory test requirements. After damage caused by the mechanical tests the package has to withstand a severe fire scenario. According to the regulations during and following the thermal test, the specimen shall not be artificially cooled and any combustion of materials of the package shall be permitted to proceed naturally.
Corresponding to results of the French institute IRSN combustion and smouldering of wood inside the impact limiter occurred during and after the fire test. An additional energy supply from a pre-damaged impact limiter to the cask could be the consequence for the safety assessment of the containment.
BAM started a first test phase to examine the issue of combustion for such kind of package components. The goal was to understand the phenomena under the consideration of relevant regulatory boundary conditions. Several metal buckets were filled with wood and equipped with thermocouples. The test specimens have been prepared with different damage arrangements to take the influence of the mechanical tests into account. This paper shows the experimental setup and the conduction of the tests. The first test shows that pre-damaged metal encapsulations can lead to smouldering of the wood and with this to a supplement energy release after the end of the 30 minute fire. BAM is in the preparation process for a second test phase. A thermal test will take place with a wood filled test specimen weighing about 2Mg.
In Germany, spent nuclear fuel and high active waste from reprocessing is stored in transport and storage containers with double lid systems that are equipped with metal seals completing the primary sealing barrier. The tasks of the Bundesanstalt für Materialforschung und -prüfung (BAM) within the interim storage licensing procedures ruled by the German Atomic Energy Act include the long-term safety evaluation of the container design regarding the permanently safe enclosure of the inventory. In order to generate a knowledge base for the safety evaluation, research regarding the long-term behavior of the critical components is performed. So far, the containers are licensed for an interim storage period of 40 years. However, due to significant delays in establishing a final repository, the required time span for interim storage is expected to increase significantly. Thus, a widespread investigation program is run to gain systematic data on the long-term behavior of the seals and to develop prediction models.
Long-term seal investigations consider the development of their restoring seal force, their useable resilience and their achievable leakage rate caused by aging at temperatures ranging from room temperature up to 150 °C. This year, the total time span of the tests reaches 10 years. Furthermore, seal segments are aged at the selected temperatures for up to 300 days. From these segments additional information on the sealing behavior, changes of the seal contact and the material behavior is gained.
This contribution deals with the current results of the long-term seal investigations at BAM. Furthermore, insights of the more in-depth component and material investigations of the metal seals with focus on the seal contact development are discussed and the ongoing work aiming for an analytical description of the thermo-mechanical aging effects on metal seals are presented.
The safety demonstrations realized by applicants in the case of approval request for the package designs containing enriched UF6 have to take into account some specific technical issues. Concerning the shipment of enriched UF6, the package designs consist in general of a filled 30B cylinder surrounded by an overpack.
The description of the content, considering the UF6 origin, i.e. natural or reprocessed, shall be clearly justified especially when the UF6 isotopic composition exceeds the limits specified in ASTM standards.
Concerning the containment of the UF6, the applicant shall demonstrate in all conditions of transport the leak-tightness of the valve and plug of the cylinders filled with enriched UF6. In this regard, when mechanical justifications are based on numerical calculations, the absence of contact between these components of the cylinder and the internal surfaces of the overpack after the regulatory drop tests shall be shown to respect the IAEA regulations.
Furthermore, the representativeness of the ballast used to simulate the behaviour of the UF6 loaded within the cylinder shall be justified if drop tests are performed. The representativeness of the ballast should also be justified for numerical calculations.
In addition, the applicant shall demonstrate that the melting temperature of the valve and the plug, including the tinned joint, will not be exceeded during the regulatory fire test.
Finally, specific provisions relative to the use of plugs and the maintenance of cylinders should be included in the safety analysis report.
A Memorandum of Understanding (MOU) between Argonne National Laboratory (Argonne) and the German Bundesanstalt für Materialforschung und -prüfung (BAM) was signed in October 2014. Its objectives are to promote cooperation among scientists and specialists at Argonne and BAM and establish a framework for collaboration in advanced technologies associated with the back end of the nuclear fuel cycle. Collaborative activities involving Argonne and BAM may be implemented through the promotion of joint research activities and scientific workshops and conferences; exchange of technical information; and visits by scientists, specialists, and graduate, postgraduate, and Ph.D. students. The MOU was renewed in January 2020 for another five years. Highlights of the progress and achievements are provided in identified topical areas for collaboration on ageing management guidance, storage, transportation and disposal R&D, joint conference activities, and conclude with the benefits of the MOU in fostering Argonne and BAM collaboration.
For stowage and retention during transport, the International Atomic EnergyAgency (IAEA) Transport regulations (SSR-6) only require that the package shall be securely stowed. Additional information is given in the IAEA guidance material (SSG-26), that includes specific load factors that should be considered in the safety demonstrations.
Nevertheless, applicants commonly use other load factors to justify the design of the packaging attachment points. In particular, acceleration values vary between different countries and this may lead to difficulties during the validation of foreign approval certificates.
Discussions with applicants identified that the load factors presented in the IAEA guidance material are not necessarily relevant for the different transport modes. For some modes, the load factors are representative of situations occurring in routine conditions of transport. For other modes, the load factors are more representative of situations occurring in normal or accident conditions of transport.
Furthermore, the origins of the specified values are not clear. In this context, some IAEA TRANSSC members decided to review the stowage guidance. An international working group was constituted in 2013, including representatives of competent authorities, technical support organizations, and transport stakeholders. Several topics were discussed
and many questions were raised during the two year review. For example, some discussions focused on Paper No.1031 2 the conditions which have to be considered for stowage design, both as relevant to the load factors used for strength and fatigue analysis, as well as the criteria which have to be considered for the package attachment points. In addition, related questions on operational aspects were also discussed. Overall, the working group concluded that new guidance material was warranted as input into SSG-26.
As a result of those discussions, the international working group produced new guidance material for stowage in transport, addressing each of these topics. The proposal to modify the IAEA guidance material (SSG-26, Appendix IV), presented by France to the TRANSSC in the 2015 initiated Review cycle of the Regulations, was unanimously accepted and will be implemented in the next edition of the guidance material.
For the long-term storage of spent nuclear fuel and other high-Level radioactive waste, dual purpose casks for the transportation and storage are widely used. Usually, these casks consist of a thick-walled monolithic cask body and a bolted double barrier lid System equipped with metal seals (consisting of an inner helical spring and two outer metal layers with a total cross-section diameter between 5 and 10 mm) to ensure the long-term safe confinement of the radioactive inventory.
Based on analytical and experimental studies, such metallic Systems have been qualified of storage periods of up to four decades so far as long as proper manufacturing and assembling conditions are met. More than 20 years of Operation experience with a permanently increasing number of casks has shown full compliance with the expectations.
In the meantime many countries have to face a significant delay of establishing a high-level radioactive waste repository what leads to Need for extending interim storage periods for probably 60, 80 or 100 years.
For that reason the long-term performance of metal seals for such time periods is of specific internst to demonstrate their sufficient safety function and specific investigations have already been launched, e.g. by BAM, to gain more experimental data and to develop validated models to extrapolate the seal performance to longer periods of time and or requested temperature levels.