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Validation of numerical simulation models for transport and storage casks using drop test results
(2012)
The safety assessment of new designs for transport and storage casks for radioactive materials is a challenging task accomplished using different methods such as prototype tests, model tests, calculations and analogy reflections. At BAM (Federal Institute for Materials Research and Testing), the test procedures for the mechanical IAEA (International Atomic Energy Agency) test conditions often start with preliminary finite element (FE) calculations mostly with a small-scale cask model for verification of the proposed test cask instrumentation and test plan. On that basis the extensive test cask instrumentation is applied and checked. After that, a series of drop tests consisting of different test sequences is performed.
Following the drop tests, numerical post-analyses are carried out. These analyses offer the possibility of a detailed calculation and assessment of stresses and strains in the entire test cask construction. The calculation results have to be carefully compared with the measurement data over the impact history to find out all relevant parameters for a realistic simulation of the impact scenario. The desired ideal boundary test conditions often cannot be met exactly during the drop tests. Therefore, the numerical post-analyses are carried out by using the real boundary conditions of the drop tests. The objective is to find a validated model, where the results of the numerical simulations satisfactorily meet the experimental results.
Under test conditions according to the IAEA transport regulations, casks are usually equipped with impact limiters and dropped onto a so-called unyielding target. In general, it is difficult to verify a complex FE model by using results from only one drop test because of the complex impact process and the complex structure of such packages. After each drop test, numerical post-analyses should be carried out. Only if all drop tests were simulated successfully by using the same FE model under different test conditions, it is possible to obtain a validated numerical model for further investigations. In this case the results of the numerical simulations meet satisfactorily the experimental results. In this paper a study is presented, where the influence of different components on the cask loading is investigated systematically.
Rubbers are widely used as main sealing materials for containers for low and intermediate level radioactive waste and as additional component to metal seals in spent fuel and high active waste containers. According to appropriate guidelines and regulations safe enclosure of the radioactive container contents has to be guaranteed for long storage periods as well as down to temperatures of -40 °C for transportation. Therefore the understanding of seal behavior is of high importance.
In this paper we focus on the behavior of elastomer seals at low temperatures with regard to poten-tial decrease of leak-tightness. In addition, changes in material properties due to aging effects over long periods of time and their influence on the seal performance is investigated.
It is known that material properties of rubbers are strongly temperature dependent. At low tempera-tures this is caused by the rubber-glass transition (abbr. glass transition). During continuous cool-ing, the material changes from rubber-like entropy-elastic to stiff energy-elastic behavior, that al-lows nearly no strain or retraction due to the glass transition. Hence rubbers are normally used above their glass transition but the minimum working temperature limit is not defined precisely.
Aging of elastomer seals is important, as possible dynamic loads may have to be considered during the whole interim storage period (so far approved in Germany for up to 40 years) and for transporta-tion after storage.
For the investigations, fluorocarbon (FKM) and ethylene-propylene-diene (EPDM) rubbers were selected as they are often used in radioactive waste containers. Some materials were purchased from a commercial seal producer and some materials were compounded and cured at BAM. The elastom-ers where studied by several thermo-analytical methods and compression set to characterize the material behavior at low temperatures. Additionally component tests were performed to determine the breakdown temperature of the sealing function of complete elastomer O-rings.
German approach and experience feedback of transport ability of SNF packages after interim storage
(2013)
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 (ESK).
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 dependant 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 behavior of components and specific operating procedures of the package.
Present research and knowledge concerning the long term behavior 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 behavior 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.
Existing spent nuclear fuel (SF) and high active waste (HAW) management policies and practices worldwide are the result of past presumptions that sufficient reprocessing and/or disposal capacity would be available in the near term. Consequently, in the past many countries have developed specific solutions for different periods of time due to their individual national nuclear policies.
In Germany the concept of dry interim storage in dual purpose metal casks before disposal is being pursued for SF and HAW management and transport and storage licenses have been issued accordingly. The current operation licenses for existing storage facilities have been granted for a storage period of up to 40 years. This concept has demonstrated its suitability for over 20 years so far. Relevant safety requirements have been assessed for the short-term as well as for the long-term for site-specific operational and accidental storage conditions. But in the meantime significant delays in the national repository siting procedure occurred which will make extended storage periods necessary in the future.
Up to the end of this decade, corresponding to the planned date of starting operation in the final disposal KONRAD for non-heat generating waste in Germany, a lot of efforts are needed to condition and package the radioactive waste in containers certified by BfS (Federal Institute for Radiation Protection). This waste is produced by public sector and industry as well as nuclear energy facilities, which result in more than a half of the actually declared quantity growing especially after the phase out decision of nuclear power production in Germany and the subsequent decommissioning of nuclear power plants. BAM (Federal Institute for Materials Research and Testing) acts as responsible authority on behalf of BfS for design testing under consideration of the KONRAD requirements. Within the assessment procedure of containers, BAM has to verify the application documents, including material qualification, container geometry, corrosion protection, leakage rate and operational and accidental loading, and to evaluate quality assurance measures. Besides the previous completed approvals for various containers such as steel sheet and cast iron box-shaped containers and concrete cylindrical ones, BAM is actually carrying assessments for other types such as cast iron cylindrical containers and “old” steel sheet box shaped ones. The so called “old” containers present already loaded containers without any KONRAD certification, currently stored at licensed interim storages. In the benefit of the container assessment, BAM operates design testing facilities for drop and fire tests which are also used for research objectives to improve and expand the evaluation methods such as research project ConDrop. The latter deals with numerical simulations and analyses tools for further precise predictions about unfavorable drop test scenarios, safety margins, and design sensitivities for steel sheet containers. Furthermore, during the assessment procedure, several specifications about the containers, the inventory to be disposed and the required safety level have been intensely debated by authorities and applicants. Based on its experience in qualifying containers, BAM has been commissioned to identify insufficiently specified aspects in the waste acceptance criteria and to propose clearer definitions with regard to the secondary regulations in the planning approval notice. This paper focuses on aspects of the contribution of BAM in certifying KONRAD waste containers. This will include the current state of design testing procedures, evaluation methods that are available or are being developed and the main topics within the KONRAD requirements being actually specified by BAM.
In Germany dual purpose casks for spent nuclear fuel (SF) or high level radioactive waste (HLW) are used for safe transportation and interim storage. Key safety issues in both fields are the safe enclosure of the radioactive material, the safe removal of decay heat, securing nuclear criticality safety, limitation of radiation exposure to acceptable levels and keeping it as low as reasonable achievable (ALARA principle).
Whereas these safety requirements during transportation are ensured by the Type-B(U)-design testing and approval procedure on basis of internationally agreed IAEA (International Atomic Energy Agency) requirements storage licenses are issued on national and site specific safety analyses.
This paper presents experiences from recent interim storage safety evaluation and licensing procedures in Germany on basis of a Type-B(U) certified cask designs concerning specific differences from Operation and accident conditions inside the storage facility. The focus is laid on the interaction of cask and building structures with regard to shielding, heat removal and accident analyses including aircraft crash. Basic safety is assured by thick-walled metal casks with monitored double lid Systems. They also reduce radiation to levels where workers can safely operate and maintain the casks inside the storage facility. The storage building e. g. provides additional shielding and protection against extemal hazards depending on the building construction. In addition, the storage building helps to reduce radiation levels at the boundary of the storage site well below regulatory limits given by the German Radiation Protection Ordinance (StrSchV).
This publication summarizes the work carried out during an IAEA coordinated research project (CRP) on safe and reliabe management of spent nuclear fuel. This work contributes to the overall goal of demonstrating the performance of spent nuclear fuel and related storage systems components over long durations, and facilitates the transfer of this knowledge to Member States. The technical areas addressed by the CRP participants were related to potential degradation mechanisms in metal casks and concrete overpacks, such as stress corrosion cracking; long term integrity and performance of the fuel cladding; thermo-mechanical behaviour of the metal seals; and long term gamma and neutron shielding capability. The participants also considered past and ongoing demonstration programmes. For each technical topic addressed the main conclusions have been drawn, including information on the development of specific monitoring and inspection techniques as well as future opportunities for closing relevant data gaps. A major achievement is also the establishment of a worldwide network of experts working on current research projects to demonstrate the long term performance of spent fuel in dry storage systems.
The evaluation of cladding integrity is a major issue to be demonstrated in Germany for extended interim storage periods up to 100 years and subsequent transportation considering operational and accidental conditions with respect to reactor operation, cask drying and dry interim storage. The chemical reaction between the zirconium fuel cladding and the cooling water in water-cooled reactors produces hydrogen and zirconium oxide. Hydrogen diffuses into the cladding and precipitates as zirconium hydrides when the solubility limit is reached, preferably oriented in hoop direction. At high temperatures during vacuum drying procedures, the hydrides can dissolve. Over a succeeding period of slow cooling with existing hoop stress the hydrides precipitate again, but partly reoriented along the radial direction of the cladding. This change of microstructure in combination with a decreasing temperature (0.5...2 K/year) during (extended) interim storage and additional mechanical load by handling procedures or under accident conditions could lead to a potential cladding embrittlement and consequently increased failure probability. The current research project BRUZL (Fracture mechanical analysis of spent fuel claddings under long-term dry interim storage conditions) has been launched by BAM to investigate potential sudden brittle failure of spent fuel claddings at small deformation under long-term dry interim storage conditions and is presented.
The safety evaluation of cask components made of ductile cast iron includes investigations to prevent brittle fracture. Generally, ductile cast iron is endangered by brittle fracture especially at low temperatures (down to -40°C) and in combination with existing crack-like material defects. An applicable method is the assessment of fracture resistance using fracture mechanics according to the IAEA guidelines. The approach is based on the prevention of fracture initiation. For application of these principles for drop loads, account must be taken both of dynamic stresses within the component and dynamic material behavior. Basically, the dynamic stress intensity factor of postulated pre-existing crack-like defects is compared with the dynamic fracture toughness of the material. Applicable numerical and experimental methods for the safety assessment of cask components are demonstrated for the case of an artificially pre-cracked cylindrical cast iron cask which undergoes dynamic loading conditions as result of the hard impact between the cask and a concrete target. The proposed evaluation procedure is a combination of numerical and experimental steps. Exemplarily, the calculated stress intensity factor is compared with measured fracture toughness values from single edge notched bending specimens.
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 investigation of the long-term performance of sealing systems employed in containers of radioactive waste and spent nuclear fuel is one research focus area for division 3.4 \Safety of Storage Containers" at the Bundesanstalt für Materialforschung und -prüfung. The behaviour of metallic seals, which are employed in interim storage casks for spent nuclear fuel and high active waste, over time is of high importance for ensuring safe enclosure. Therefore, investigations on these systems were started at BAM to get a general understanding of the relevant processes. Our investigations comprise investigations on different parameters which influence the seal Performance and the main part is focussed on the time and temperature dependent behaviour. Our aim is to understand the longterm behaviour of the sealing systems for evaluation of their performance during possible extended interim storage and subsequent transportation. In this contribution an overview of the performed tests and their respective results is given. A focus lies on a comparison between seals with large torus diameter in comparison to small diameters.
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.
Beginning in 1976 the former iron ore mine Konrad was geologically investigated concerning its suitability as a location for a German final repository for low and intermediate level waste. After a comprehensive licensing procedure it was approved and is now being prepared and equipped for the planned start of storage in 2027. At the end of the 1980s, the requirements for waste containers for storage in this repository were defined for the first time on the basis of an incident analysis of the on-site handling procedures, beginning with the unloading of the packages after the arrival at the site up to the final positioning in the underground storage galleries. A 5-meter drop onto the rock ground of the mine was identified as the covering case for high mechanical requirements (ABK II container class). In contrast to the 9-meter drop according to the IAEA Regulations for the Safe Transport of Radioactive Material, the 5-meter drop is performed onto a hard but not essentially unyielding target. The container is typically not protected by an impact limiter. The requirements for a potential test facility are described in the regulations for the Konrad repository. The mechanical strength of the target is defined as a concrete strength equivalent to the identified properties of the rock ground of the mine.
Since 1991 BAM has consistently used precast reinforced concrete slabs as target for drop tests in the framework of licensing procedures as well as in research projects. While the original design fulfilled the requirement for the integrity of the concrete slab in most cases, it failed when drop tests onto an edge of a container were performed. A redesigned concrete slab developed in a research project and suggested as a reference target in 2009 has been successfully used in Konrad licensing procedures since then.
The paper gives a brief overview of the historical development and it describes BAM’s efforts and approaches to continuously guarantee a concrete slab of defined quality and to provide a test setup for valid drop tests from the applicants and the authorities view.
Suitable containers for receiving high-level radioactive waste are a key elements of every repository system concerning different host rocks like rock salt, claystone and crystalline rock as being considered in Germany. The disposal container design is decisive regarding boundary conditions for the transport, emplacement techniques, and the operational and long-term safety assessment. The systematic derivation of the requirements to be placed on disposal containers forms the basis for a targeted, comprehensible and transparent development of containers that meet the requirements. The presentation summarizes the outcomes of the research project KoBrA.