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Organisationseinheit der BAM
Eingeladener Vortrag
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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.
The properties of spent nuclear fuel (SNF) rods change significantly during their operation life in the reactor core. Further changes occur after their discharge mainly due to the heating-cooling processes and possible ageing associated with the cumulative effects of radioactive decay induce damage in the fuel. Such changes may affect the response of the SNF rods to mechanical solicitations corresponding to normal and accidental conditions.
Research activities at JRC-KARLSRUHE aim at assessing the integrity of SNF rods and processes which might affect their mechanical properties during their interim storage, transport or other handling operations. JRC Hot Cell facilities have been fully adapted to fulfil the experimental goals. The number of experiments that can be performed, however, is limited and there is an acute need to model them, using this process to validate codes, to deeper understand and to extend the results gained at the JRC beyond the conditions that have been tested. For the experimental campaigns two devices for gravitational impact and 3-point bending tests were developed and installed in a hot cell. Segments of real SNF rods pressurized at their original pressures after discharge have been investigated. The setup is fully operational and new results are reported continuously.
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.
As part of the evaluation of a package for the safe transport of radioactive waste the regulations of the IAEA International Atomic Energy Agency shall be fulfilled. The regulations define requirements for the package and specify mechanical and thermal test conditions. Different methods are allowed for the test performance to demonstrate compliance with the regulations. Next to calculational approaches and the use of models of an appropriate scale, the performance of full-scale testing with prototype packages respectively full-scale models is applied.
The use of full-scale models has several advantages within the complete safety assessment procedure for a transport package approval. Scaling and corresponding similarity questions don’t have to be considered, additional material investigations can be limited and analyses to transfer test results to the original package design are reduced in number and complexity. Additionally, experience for future serial design procedures can be built up during manufacturing and assembling of the test model.
BAM operates different drop and fire test facilities south of Berlin, Germany. BAM has started to perform a drop test campaign with a full-scale model of 120 metric tons weight for a transport package approval procedure. The paper describes experience with test preparation, drop performance and additional analyses. The measurement concept is explained and test goals regarding the package safety assessment and evaluation of safety margins are introduced.
Packages for the transport of radioactive materials shall fulfil the requirements of the IAEA regulations for the safe transport. The requirements define mechanical and thermal test conditions including criteria ensuring the package design’s ability to withstand severe accidents and provide a high level of technical safety. Different methods can be used for safety demonstration showing compliance with the regulations.
The central part of a safety demonstration which is presented in this paper was a comprehensive drop test program with a full-scale model of a transport package accompanied by pre- and post-test FE analyses. Using full-scale drop test models allow the benefit that similarity and scaling issues become a significant smaller issue, additional material investigations can be limited and analyses for transferring test results to the original package design are reduced. Additionally, experience for the future serial packaging manufacturing and handling procedures can be collected in a very early state of the design approval process. The pre-test finite element analyses derived and justified the drop test program consisting of several drop sequences with different drop orientations of the specimen. The performance and the results of the drop test sequences shows the manageability and the advantage e.g., in view of the direct availability of test results for the package licensing. On the other hand, the drop test performance shows the difficulties during handling and the need for additional equipment during preparation of the specimen.
The package presented was intended for the transport and storage of compacted radioactive waste from reprocessing of spent nuclear fuel assemblies - designed and applied for approval by the AGC consortium. The project ended in 2021. The package design was characterized by a cask body made of a forged thick stainless-steel shell, a bolted double lid system with metallic gaskets and wood filled shock absorbers at both ends. The total mass of the entire transport package including content was 120,000 kg, the total length was about 7000 mm and the diameter approximately
3000 mm, both measures include the shock absorbers.
The paper provides an insight into the performance of a full-scale drop testing campaign within the package safety evaluation and shows some selected test results.
TN®24 E, a new package design, was developed and applied for by Areva TN for German transport approval certificate. The certificate was issued by BfS, Federal Office for Radiation Protection, on 24th of July 2013. The package is designed as a dual purpose cask for transport and storage of up to 21 PWR spent fuel assemblies from German NPPs. BAM is the German competent authority responsible for the design assessment of RAM packages regarding mechanical and thermal safety cases, activity release analysis and all issues of quality assurance during manufacturing and operation of packages. Certain assessment experiences as well as new developments resulting from the BAM TN®24E approval procedure are presented. The mechanical safety case of the TN®24E is based mainly on finite element calculations, which were verified by the TN®81 1/3 scale drop test program performed at BAM. Thermal analyses rely upon calculations, while the activity release criterion is based upon leakage rate results of TN®81 drop tests. The BAM-GGR 012 guideline for the analysis of bolted lid and trunnion systems has been fully implemented. Due to requirements by BAM, AREVA TN developed a new assessment strategy for fracture mechanical evaluation of welding seams. The material qualification and documentation is also an important aspect of BAM assessment; the qualification of borated aluminum basket material, the determination of strength values for thermal aged hardened aluminum alloys for the basket or consideration of high burn-up fuel assemblies are remarkable issues in this context. In addition, the consideration of the material compatibility, especially taking into account a transport after 40 years of dry interim storage in German facilities, has gained significant importance in the licensing process of the TN®24E. Next to obvious mechanical issues such as the assessment of shell ovalization under 9 m drop test scenario and its impact on basket load, thermo-mechanical interactions had to be addressed in the safety case. Due to BAM requirements, AREVA TN performed a full thermo-mechanical analysis of the cask behavior under fire test conditions.
Federal Institute for Materials Research and Testing (BAM) is the responsible authority for the
assessment of mechanical and thermal safety as well as quality issues within the licensing procedure
for transport casks for radioactive materials. The assessment includes a brittle fracture analysis of all
relevant cask components.
For cask bodies where a steel bottom is welded to a steel shell, specimens for testing of sufficient
fracture resistance can normally not be extracted directly from the welding seam area of the cask.
Therefore, a methodology has to be developed to ensure compliance with the safety analysis
considerations and the approved design specification. This paper describes such a methodology
including brittle fracture analysis and de-termination of required material properties as well as aspects
of quality assurance during manufacturing.
Germany had been transporting spent fuel to the reprocessing plant in Sellafield and La Hague for decades until around 2005. Resulting from the obligation to take back the vitrified high-level waste from reprocessing six CASTOR® HAW28M filled with 168 canisters with radioactive waste in vitrified form were transported from Sellafield to the interim storage facility Biblis in 2020. Rail wagons were used for the transport to the port in Barrow-in-Furness, where they were loaded into a dedicated seagoing vessel, certified as INF Class 3 according to the INF Code. This was the first time that vitrified high level waste with considerable heat load was transported under a German design approval certificate.
BAM was involved in the authority assessment of the conditions for the sea transport. For the first transport BAM required among others, assessment of temperature distribution during transport, logging of temperatures of cargo bays and graphical imaging of temperatures of the bay with the cask in order to ensure compliance with temperature specifications, e.g. maximal neutron absorber and gasket temperatures. Special interest was taken in the identification of possible events exceeding the specified temperatures considering the different philosophies of IMDG code and its supplement INF code regarding temperature control of hatches. Results show compliance with assumed conditions.
Packages for intermediate level waste (ILW) often contain residual water besides the actual waste. The water either exists as obvious free water or it may be bound physically or chemically, e.g. as pore water. A water driven gas generation could occur by vaporisation and by radiolysis. Steam as the result of vaporisation causes an increasing pressure inside a package and can affect corrosion. Vaporisation and condensation processes itself change the thermal behavior of the content especially during strongly unsteady thermal situations like accident fire situations. Radiolysis changes the chemical composition of the content which could cause an unexpected interaction, e.g. hydrogen embrittlement. Besides the pressure build-up the radiolysis of water generates hydrogen and oxygen, which can be highly flammable respectively explosive. The gas generation caused by vaporisation and radiolysis must be taken into account during the design and the safety assessment of a package. Pressure build-up, a changed thermal behavior and content chemistry, and especially the risk of accumulation of combustible gases exceeding the limiting concentration for flammability has to be considered in the safety assessment. Approaches to ensure the transportability of stored packages due to radiolysis will be discussed.
This paper presents data obtained from experiments performed using a bending test set-up developed at the Joint Research Centre (JRC) – Karlsruhe, for spent fuel segment testing. Adjustable sample holders, loading modes and other experimental conditions can be im- plemented in the experiments to study the effects of different deformation ranges up to cladding failure. The experimental set-up has been adapted to hot cell remote controlling and has a modular configuration, which allows manual and motor-driven loading option. The device has been calibrated on hydrogenated, unirradiated cladding tube segments filled with alumina pellets. The final application of present set-up is to test non-defueled spent fuel rod segments, pressurized to the original spent fuel rod pressure level. The range of applicability of this device, the scope of the experimental program and the first results from actual bending tests will be discussed.