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This paper describes the analytical, numerical and experimental investigations on the phenomenon of
interactions between cask components and content of packages for the transport of radioactive material during drop
tests required according to the IAEA Regulations for the Safe Transport of Radioactive Material.
Radial and axial gaps between cask components and content are usually necessary for thermal reasons but larger
gaps can exist because of the geometrical dimensions of the specified content. Consequently interactions between
content and cask components (lid system, cask body, etc.) are possible and can not be excluded during drop tests.
Interactions in this context are relative movements between cask and content which are mainly due to elastic spring
effects after releasing the cask for the free drop. These relative movements can cause interior collisions between
content and cask during the main impact of the package onto the unyielding target.
Drop tests with various types of Type A and Type B packages fully instrumented with strain gauges and
accelerometers showed that these interactions respectively interior collisions can be considerable relating to high forces
acting on cask lids, lid bolts and the content. Of course the real quantitative consequences of the interactions depend
upon different conditions, among others the drop orientation, the design characteristics of the impact limiters, the
dimensions of the gaps, the material characteristics of the contents, etc. .
In order to investigate more precisely the phenomenon of interactions BAM carried out finite element
calculations for the named casks using the ABAQUS/ Standard and ABAQUS/ Explicit computer code comparing them
with results obtained from experiments. Additionally, tests with a simplified model instrumented with accelerometers
were carried out accompanied by finite element calculations and analytical calculations using MATHEMATICA.
The investigations on the mentioned phenomena of interaction show that they should be considered in approval design
tests and/ or calculations.
Tests with different Type B casks confirm the existence of effects with stress peaks in cask components due to
interactions between the cask and its contents. These effects can be caused by a delayed strike of the content onto surrounding cask components which cannot be excluded if the content is movable. Some results of the drop tests with two different Type B casks and with a model designed for the study of this problem are presented in the paper. Results of calculations performed with the ABAQUS computer code and by use of analytical methods to simulate the measured effects are discussed.
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.
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.
Methodological aspects for finite element modeling of lid systems for type B(U) transport packages
(2012)
The regulatory compliance of the containment system is of essential importance for the assessment
process of Type B(U) transport packages. The requirements of the IAEA safety standards for
transport conditions implies high loading on the containment system. The integrity of the
containment system has to be ensured in mechanical and thermal tests.
The containment system of German SNF and HLW transport packages usually includes bolted lids
with metal gaskets. The finite element (FE) method is recommended for the analysis of lid systems
according to the guideline BAM-GGR 012 for assessment of bolted lid and trunnion systems. 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 Type B(U) packages for
the transport of SNF and HLW. The work is based on the experiences of BAM Federal Institute for
Materials Research and Testing as the German competent authority for the mechanical design
assessment of such kind of packages. The issues considered include modeling strategies, analysis
techniques and the interpretation of results. A particular focus of this paper is on the evaluation of
the results with regard to FE accuracy, influence of the FE contact formulation and FE modeling
techniques to take the metallic gasket into account.
Packages for the transport of radioactive material are generally equipped with specific
structures (basket) to support the radioactive content in defined position. The safety function
of the basket depends on the kind of transported inventory. In case of transport cask for spent
fuel, the basket design has to ensure the subcriticality of the fissile material in all conditions
of transport in particular. Therefore the evaluation of structural integrity and neutron
absorption capability of the basket is an important part of complete safety analysis. Sufficient
heat transfer to maintain fuel assembly and cask temperature within allowable limits has to be
verified as well. Corrosion resistance is an additional requirement on basket materials owing
to contact with water during loading and unloading operations.
Computational and experimental methods or their combination along with additional material
and component tests can be used to analyse the mechanical and thermal basket behaviour
under transport conditions defined in IAEA regulations. By deciding between the analysis
methods, the design features (including material selection concept) as well as specific safety
function should be accounted.
In approval procedures of transport packages for radioactive materials, the competent
authority mechanical and thermal safety assessment is carried out in Germany by BAM. Some
questions of safety evaluation of basket designs are discussed in this paper based on the BAM
experience within approval procedures. The paper focuses primarily on the mechanical
behaviour of baskets with regard to the assumptions that have to be used in the criticality
safety demonstration. The state of the art methodologies for computational basket stress and
deformation analysis as well as for interpretation of drop tests results are presented.
The Federal Institute for Materials Research and Testing (BAM) intends to provide a new guideline for the analysis and assessment of lid and trunnion systems. This guideline is discussed in the present paper with respect to its contents and structure, its compatibility with other guidelines and selective technical issues to demonstrate its use. Particularly, the motivation, the principles and the background of the guideline are addressed. Therefore, the present paper discusses current conclusions drawn from experimental results and the wide use of advanced analysis methods such as finite element analysis as well as the incorporation of these aspects in the guideline draft.
Dynamic and quasi-static FE calculation of impact from 9 metre drop of spent fuel transport cask
(2009)
The drop from 9 m height onto a rigid target is one of the required proofs of safety for packages of radioactive materials. Direct dynamic finite element method (FEM) calculations and combined two step analytical quasi-static finite element (FE) calculations are both applied for the simulation of a 9 m drop test. This paper gives a comparative example of both approaches for the 9 m side drop simulation of the spent fuel transport cask with wood filled impact limiter. The model for dynamic FEM consists of cask body and impact limiter. Detailed material properties and geometry descriptions from each component of the impact limiter are required. The results (stress fields in the cask body) are obtained directly from the calculation. The combined method provides as intermediate results the force-deformation characteristic of impact limiter. The maximum impact limiter force determined by the law of energy conservation during the drop is then - in a second step - applied on the cask body in a quasi-static FE model in order to calculate the stresses. In this paper, the rigid body deceleration and the maximum stress in the middle of the cask body are used for the comparison between the dynamic FEM and the combined method. Similar maximum rigid body deceleration-time curves were obtained by both methods for the horizontal 9 m free fall. Concerning the stress in the cask body the dynamic FEM results oscillate about values calculated by the combined quasi-static approach. If the combined quasi-static approach is used in the safety assessment of a cask, a suitable factor has to be applied on its results to take into account the additional dynamic effects.
The most relevant source of reference for load cases of packages for the transport of radioactive material under routine conditions is Appendix IV of the International Atomic Energy Agency (IAEA) Advisory Material TS-G-1·1. The acceleration values given there leave many questions open and have to be agreed upon with the involved competent authorities. Consequently, the actual load cases applied for a safety analysis may differ widely and could cause problems for international transport. To avoid such difficulties, it seems obvious that the according passages in TS-G-1·1 should be modified with the objective of making the load case data more consistent for a harmonized application in the IAEA member states. Papers pointing out in this direction were discussed at the PATRAM 2010 conference. The way to improve the acceleration data may lie in investigations and measurements considering, for example different types of vehicles and package masses. However, what should be the goal of such kind of surveys? Can such investigation provide values for every worldwide load case possible under routine conditions of transport? In particular, the different designs of vehicles give a reason to doubt that such an aim is realistic. The approach in this paper is to show that a less ambitious aim is more effective. Therefore, an approximate scheme is preferred, which renders better assistance in determining the appropriate acceleration values.
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.
This paper demonstrates exemplarily how numerical and experimental approaches can be combined reasonably in mechanical assessment of package integrity according to the IAEA regulations. The paper also concentrates on the question about how static mechanical approaches can be applied, and what their problems are in relation to dynamic calculation approaches. Under defined impact tests, which represent accident transport conditions, the package has to withstand impact loading, e.g. resulting from a 9 m free drop onto an unyielding target in sequence with a 1 m puncture drop test. Owing to the local character of the interaction between the puncture bar and the cask body, it is possible to develop a dynamic numerical model for the 1 m puncture drop which allows an appropriate simulation of the interaction area. Results from existing experimental drop tests with prototype or small scale cask models can be used for verification and validation of applied analysis codes and models. The link between analysis and experimental drop testing is described exemplarily by considering a regulatory 1 m puncture bar drop test onto the cask body of a recently approved German high level waste transport package. For the 9 m drop test of the package, it is difficult to develop a dynamic numerical model of the package due to the complexity of the interaction between cask body, impact limiters and unyielding target. Dynamic calculations require an extensive verification with experimental results. The simulation of a 9 m drop of a package with impact limiters is thereby often more complex than the simulation of a 1 m puncture drop onto the cask body. A different approximation method can be applied for the consideration of dynamic effects on the impact loading of the package. In a first step, maximum impact force and rigid body deceleration of the cask body during the impact process can be calculated with simplified numerical tools. This rigid body deceleration can subsequently be applied on a verified static numerical model. Dynamic effects, which cannot be covered by the static numerical analysis, have therefore to be considered by using an additional dynamic factor. The paper describes this approach exemplarily for a 9 m horizontal drop of a typical spent fuel cask design.
Methodological aspects for finite element modelling of lid systems for type B(U) transport packages
(2012)
The regulatory compliance of the containment system is of essential importance for the assessment process of Type B(U) transport packages. The requirements of the International Atomic Energy Agency safety standards for transport conditions imply high loading on the containment system. The integrity of the containment system has to be ensured in mechanical and thermal tests. The containment system of German spent nuclear fuel and high level waste transport packages usually includes bolted lids with metal gaskets. The finite element (FE) method is recommended for the analysis of lid systems according to the guideline BAM-GGR 012 for the assessment of bolted lid and trunnion systems. The 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 type B(U) packages for the transport of spent nuclear fuel and high level waste. The work is based on the experiences of the BAM Federal Institute for Materials Research and Testing as the German competent authority for the mechanical design assessment of such kind of packages. The issues considered include modelling strategies, analysis techniques and interpretation of results. A particular focus of this paper is on the evaluation of the results with regard to FE accuracy, influence of the FE contact formulation and FE modelling techniques to take the metallic gasket into account.
Methodological aspects for numerical analysis of lid systems for SNF and HLW transport packages
(2013)
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 BAM-GGR 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.
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 interaction between the package lid system and internal Content during mechanical drop testing is a decisive matter in evaluating Impact loads and the safety of the package. In the case of movable contents ist impact onto the inner side of the package lid can cause additional load peaks on the lid and the lid bolts. Some aspects of this issue were discussed on the basis of experimental results from instrumented drop tests with transport casks and on the basis of analytical approaches.
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.
Acceptable limits for activity release from transport casks for high-level radioactive material specified in the IAEA regulations must be kept by the integrity of cask body and the cask sealing system. BAM as the German competent authority for mechanical, thermal and containment assessment of packages liable for approval verifies the activity release compliance with the regulatory limits. The specification of conservative package design leakage rates is one of the most important aspects in assessment.
Metal seals of the Helicoflex® Type are usually used to ensure the required package tightness for both, storage, and transport of the cask before and after storage. Due to the long-term use the seal behavior is influenced by temperature and time. The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport specified in the regulations can have a significant effect on the leak tightness of the sealing system. Whereas the safety for application of new, non- aged Helicoflex® seals is verified sufficiently, there are still technical data gaps concerning the efficiency of aged Helicoflex® seals.
BAM performed experiments to learn more about the sealing efficiency of aged Helicoflex® seals with Aluminum and Silver outer jackets. The seals were compressed in test-flanges and for artificial ageing the complete flange systems were stored in an oven for several month at a high temperature. After this ageing procedure the flanges were opened completely, the seals were moved a little to vary the contact area, and the flanges were compressed again. During the compression and decompression tests after the aging, load-deformation characteristics of the seals, and leakage rates were measured.
With these tests a load situation was simulated, which can occur in the regulatory drop test of the cask: Under high impact loads the bolted lid can lift a little for a short moment, allowing a little movement of the seal, so that the contact area can change before compressing again.
Details about test conditions and first results will be presented in the poster.
Experimental and numerical investigation of prestressed bolt connections under lateral displacements
(2022)
The containment system of transport packages for spent nuclear fuel and high-level waste usually includes bolted lids with metal gaskets. The packages are assessed to specific transport conditions which are specified in the IAEA safety standards SSR-6 (IAEA 2018). These transport conditions, especially the so-called accident conditions of transport, imply high dynamic loading on the lids and the bolt connections of the package. The response of the lid systems on the mechanical accident conditions is generally investigated by experimental drop tests or numerically, e.g., by finite element analyses. The interpretation of the drop test results for the verification of the numerical models is often not obvious due to the complex superposition of different effects in the real tests. BAM has started a research project to get a better understanding about the behavior of prestressed bolt connections under loadings typical for these drop tests. In this context an experimental test set-up was developed to investigate the response of a single bolt connection under a prescribed lateral displacement of clamped parts. The bolt is instrumented by strain gauges to get the pretensional, the torsional and the bending stress in the bolt shank. Furthermore, the lateral movement and the tilt of the bolt head is measured during the test. A finite element model of the test set-up has been created in Abaqus FEA (Simulia 2021). The very detailed instrumentation of the test set-up shall give the opportunity to investigate and validate the numerical model. The aim of this paper is to give an overview about the proposed research project and to present first results.
Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the regulations of the International Atomic Energy Agency. 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 capabilities under These conditions. Cracks or failures in the fuel rod cladding can cause the release of gas, volatiles or fuel particles into the cavity. The amount of substances in the cavity has to be considered in the assessment of the activity release and criticality safety.
The mechanical analysis of the compound system formed by the fuel rod cladding and the spent fuel pellets is very difficult due to the limited knowledge of the material properties and the insufficient understanding of the interaction between pellets and cladding and between adjacent pellets. The variation of fuel assembly properties regarding cladding material, burn-up and the history of usage makes reliable predictions of the fuel rod behavior even harder.
For a better understanding about the behavior of spent fuel rods, JRC and BAM have started a joint research project. In this context, JRC has developed a test device which allows quasi-static 3-point-bending test on fuel rod segments in the hot cell. The loads are applied with respect to the boundary conditions of the activity release assessment. This paper deals with the numerical calculation of a single fuel rod segment under bending load. The aim is to identify the governing mechanical parameters by the variation of constitutive assumptions, contact conditions, inner constraints, etc. This knowledge helps for the interpretation of the experimental results. Furthermore, the improved understanding about the behavior of the cladding-pellets system will be beneficial for the assessment of spent fuel transport conditions.
Assessment experience on packages loaded with damaged spent nuclear fuel for transport after storage
(2018)
In 2017 the first German package approval certificate was issued for a dual purpose cask (DPC) design with encapsulated damaged spent nuclear fuel. At the Bundesanstalt für Materialforschung und -prüfung (BAM) a comprehensive assessment procedure was carried out with respect to the mechanical and thermal design, the containment design and quality assurance for manufacturing and operation. Main objective of this procedure was to verify the Package Design Safety Report (PDSR) fulfils the requirements according to the IAEA regulations SSR-6.
Until now only standard spent nuclear fuel assemblies were designated for interim storage and transports. Due to nuclear phase out in Germany all other kinds of SNF in particular damaged fuel has to be packed. Therefore specific requirements have to be considered in accordance with international experiences written in IAEA technical reports. In Germany damaged spent nuclear fuel (DSNF) needs a tight encapsulation with special encapsulations and clearly defined properties.
Due to the limited amount of DSNF these encapsulations are designed for storage and transport in existing packages. From the assessment experience it has been seen, corresponding PDSR need an extensively expansion to cover the design of these encapsulations and their influences on the package. Then such well-defined encapsulations can be handled like standard fuel assemblies. The main difference to standard package components is, encapsulations with permanent closure achieve their specified condition not after manufacturing but only during operation after loading and closing.
Thus specific handling instruction and test procedures are necessary especially for welding, where BAM is able to survey the quality of this first part of operation.
Recent developments in standards and IAEA guidance material for package load attachment points
(2018)
For transport package design and operation according to IAEA regulations, the package shall be securely stowed and its retention system shall be capable to withstand load conditions of routine transport. The supporting IAEA Advisory Material SSG-26 provides information how to do that.
Up to now package designers in different countries use other load factors for the design of attachment points than those specified in the IAEA guidance material. In particular the acceleration values vary between different countries and lead to difficulties during the validation of foreign approval certificates. Therefore the IAEA started a discussion process to review the existing guidance text. An international working group was constituted in 2013. Representatives came from different stakeholders, e.g. transport operators, competent authorities and modal organizations. The discussions concluded especially on the transport conditions which has to be considered for stowage design, including on the one hand the relevance of the load factors used for strength and fatigue analysis and on the other hand the criteria which have to be considered for the attachment points.
Another standard with relevance to the load attachment of packages is ISO 10276. This standard deals with trunnion design, manufacturing and operational aspects. The regular standard revision phase started in 2017. An expert group discussed new state-of-the-art technology and proposed revised text for the ISO standard for international discussion.
The paper describes relevant tie-down aspects, gives background argumentation, and tries to support harmonized application of the revised IAEA guidance material and future ISO 10276 standard.
For disposal of the research reactor of the Technical University Munich FRM II a new transport and storage cask design was under approval assessment by the German authorities on the basis of International Atomic Energy Agency (IAEA) requirements. The cask body is made of ductile cast iron and closed by two bolted lid systems with metal seals. The material of the lids is stainless steel. On each end of the cask the wood-filled impact limiters are installed to reduce impact loads to the cask under drop test conditions. In the cavity of the cask a basket for five spent fuel elements is arranged.
This design has been assessed by the Bundesanstalt für Materialforschung und -prüfung (BAM) in view to the mechanical and thermal safety analyses, the activity release approaches, and subjects of quality assurance and surveillance for manufacturing and operation of the package.
For the mechanical safety analyses of the package a combination of experimental testing and analytical/numerical calculations were applied. In total, four drop tests were carried out at the BAM large drop test facility. Two tests were carried out as a full IAEA drop test sequence consisting of a 9m drop test onto an unyielding target and a 1m puncture bar drop test. The other two drop tests were performed as single 9m drop tests and completed by additional analyses for considering the effects of an IAEA drop test sequence.
The main objectives of the drop tests were the investigation of the integrity of the package and its safety against release of radioactive material as well as the test of the fastening system of the impact limiters. Furthermore, the acceleration and strain signals measured during the tests were used for the verification of finite-element (FE) models applied in the safety analysis of the package design.
A new dual purpose cask design was developed for the safe transport and interim storage of spent fuel elements of German research reactors. In the framework of the safety assessment within the licensing procedure the Bundesanstalt für Materialforschung und –prüfung (BAM) as competent authority performed a series of drop tests according to the IAEA-Regulations [1]. The package consists of a cylindrical thick-walled ductile cast iron cask body closed by a double lid system with metallic seals. A lid and bottom sided impact limiter consisting of a wood/steel construction limit the mechanical impact loading. The full-scale test specimen was equipped with a basket and assembled with dummy-fuel elements. The package and test specimen, respectively have a total mass of approximately 24 metric tons.
The mechanical test program included three 9m free drop tests, in horizontal, vertical and oblique cask orientation onto the lid system. Additionally, a 1m-puncture drop test followed the horizontal drop test to consider an IAEA-test sequence. The horizontal and vertical drop tests were performed at a temperature of minus 40°C. During the oblique drop test the upper impact limiter was heated to +80°C. The tests were conducted onto an unyielding target, fulfilling the requirements of the IAEA regulations [1].
The test specimen was considerably instrumented with strain gauges and accelerometers. Transient strains at selected locations of the inner and outer container walls, of the primary and secondary lid, as well as of the corresponding lid bolts were measured during the drop tests. Furthermore, decelerations in different locations at the cask body and the lids were measured. The complex geometrical deformation of the impact limiters due to the impact were determined by optical 3d- measurements using the projected fringe method in combination with multi-image photogrammetry. Before and after the drop tests the leakage rate of the lid system was determined by helium leakage testing.
The experimental results contribute to the evaluation of the package response to mechanical tests, demonstrating safety under normal and accident conditions of transport. Especially to the verification of the dynamic finiteelement model of the package used in the package design safety report [2]. The paper describes the performance of the drop tests, selected test results focusing on the lid screws and the cask body and the deformation of the impact limiters as well as impact kinematics, respectively.
A new dual purpose cask design was developed for the safe transport and interim storage of spent fuel elements of German research reactors. In the framework of the safety assessment within the package approval procedure the Bundesanstalt für Materialforschung und –prüfung (BAM) as competent authority performed a series of drop tests according with the IAEA Transport Regulations. The package consists of a cylindrical thick-walled ductile cast iron cask body closed by a bolted lid system with metallic seals. A lid and bottom sided impact limiter consisting of a wood/steel construction limit the mechanical impact loading. The full-scale test specimen was equipped with a basket and assembled with dummy-fuel elements. The package and test specimen, respectively have a total mass of approximately 24 metric tons. The mechanical drop test program included three 9m free drop tests, in horizontal, vertical and oblique cask orientation onto the lid system. Additionally, a 1m-puncture bar drop test followed the horizontal drop test to consider an IAEA-drop test sequence. The horizontal and vertical drop tests were performed at a temperature of minus 40°C. During the oblique drop test the upper impact limiter was heated up to +80°C. The tests were conducted onto an unyielding target, fulfilling the IAEA requirements. The test specimen was considerably instrumented with strain gauges and accelerometers. Transient strains at selected locations of the inner and outer container walls, of the primary and secondary lid, as well as of the corresponding lid bolts were measured during the drop tests. Furthermore, decelerations in different locations at the cask body and the lids were measured. The complex geometrical deformation of the impact limiters due to the impact were determined by optical 3D- measurements. Before and after the drop tests the leakage rate of the lid system was determined by helium leakage testing. The experimental results contribute to the evaluation of the package response to mechanical tests, demonstrating safety under normal and accident conditions of transport and especially to the verification of the dynamic finite-element model of the package used in the package design safety report.
Type B(U) packages for the transport of radioactive material have to withstand accident conditions of transport defined in the regulations of the IAEA International Atomic Energy Agency in form of different mechanical (drop) tests with a subsequent thermal test. According to the regulatory requirements the orientation of the package in drop tests shall be such to cause the most damaged state in the components performing the safety functions.
For the package lid system a 9-m drop onto the unyielding target with lid side downwards is often the most damaging orientation. The impact loads acting on the lid in this orientation result mainly from interaction between lid and internal content. In case of a movable content its impact onto the inner side of the lid can cause additional load peaks on the lid and the lid bolts. The intensity of the internal collision depends on the position of content relating to lid at the time of package first contact with target. Due to physical limitations an axial gap, which could be set in “pre-drop” configuration of package or which could spontaneously appear during the drop test, usually does not cover the maximum size possible in specific package design. In this context, the combination of drop tests with post-test analysis can be helpful to better estimate the effect of internal impact.
The paper summarized some aspects of this issue based on the BAM experience in the design assessment of Type B(U) transport packages. Additionally the paper shall support applicants in German approval procedures to reduce rounds of questions and ensure delivery of reliable safety case documents to the authorities. International discussions of this issue at the IAEA and a joint proposal by France and Germany to improve Advisory Material text will be introduced.
Internal cask content collisions during drop test of transport casks for radioactive materials
(2013)
In transport casks for radioactive materials, significantly large axial and radial gaps between cask and internal content are often present because of certain specific geometrical dimensions of the content (e.g. spent fuel elements) or thermal reasons. The possibility of inner relative movement between content and cask will increase if the content is not fixed. During drop testing, these movements can lead to internal cask content collisions, causing significantly high loads on the cask components and the content itself. Especially in vertical drop test orientations onto a lid side of the cask, an internal collision induced by a delayed impact of the content onto the inner side of the lid can cause high stress peaks in the lid and the lid bolts with the risk of component failure as well as impairment of the leak tightness of the closure system. This paper reflects causes and effects of the phenomenon of internal impact on the basis of experimental results obtained from instrumented drop tests with transport casks and on the basis of analytical approaches. Furthermore, the paper concludes the importance of consideration of possible cask content collisions in the safety analysis of transport casks for radioactive materials under accident conditions of transport.
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.
In Germany current package design safety cases include more and more advanced numerical methods, e. g. finite element analysis (FEA), often in combination with local concepts of strength evaluation of the structure. This approach requires extensive modeling and verification procedures. As a consequence the efforts of authority assessment of design safety analysis increase as well. Only the check of pre- and post-data of numerical calculations is often not sufficient for the safety assessment. On the other hand own analyses of the mechanical problem by performing an Independent numerical modeling and analyzing is not always realizable. Therefore it is necessary to look for optimized procedures of the assessment, without loss of safety. This paper shows possibilities for the assessment strategy of numerical analyses with focus on simple analytical approaches as comparative calculations. Such approaches can be helpful to support evaluation of numerical calculations in the whole assessment procedure.
Three examples are considered to Show which possibilities and limits exist to support the assessment of numerical analyses using analytical comparative calculations. Two examples of bolt and lid analysis show the influence of component and boundary stiffness on the results. Thickness to length/width ratios are partially exceeded and only fixed or free boundary conditions can be analyzed analytical. Nevertheless These analytical approaches can help to evaluate the numerical results for the assessment.
The example of a trunnion demonstrates the limits of analytical approaches. The trunnion shows a complex deformation behavior and local stresses. A single basic theory isn’t matching and a construct of several approaches is not useable for calculations of local stresses. Therefore numerical calculations during assessment are necessary.
Analytical approaches are not always purposeful but often effective to reduce the effort of assessment for numerical analysis of complex and safety relevant components of RAM packages.
Load attachment points, consisting of the lifting component and corresponding bolt connections, and lid systems, consisting of a lid, gaskets and bolt connections, are usually analyzed numerically using the Finite Element Method. Reasons for applying the FEM are complex geometry, specific load distribution and the violation of application criteria for analytical approaches. For example the elementary beam theory is not suitable for the calculation of compact trunnions and the Kirchhoff plate theory is not suitable for the calculation of lids with a large thickness to diameter ratio because the Bernoulli-hypothesis is violated. In addition for structural integrity assessments often local stress and for the evaluation of lid systems tightness local contact opening are needed, which can be obtained only through accurately discretized numerical models.
Independent comparative calculations are essential for complex calculations. Simplified but appropriate analytical approaches are an efficient way of examination. Is it possible to provide conservative analytical estimates of such kind of analysis problems? Are analytical calculations an appropriate approach in today's authority assessment business? Using the example of bolted trunnion and lid systems of a heavy package the possibilities and limits of analytical comparative calculations are shown.
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 meter drop test onto a steel bar has to be considered for hypothetical accident conditions of 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 applicant, BAM developed a complex finite element model for a dynamical ABAQUS/ExplicitTM analysis. This paper describes in detail the use of the finite element (FE) method for modeling the puncture drop test within an actual assessment strategy.
At first investigations of the behaviour of the steel bar are carried out. Different friction coefficients and the material law of the bar are 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 is developed. Strain verification is possible by results of the strain gauges located at the relevant points of the cask model. The influence of the finite element meshing is described.
Finally, the verified FE half-scale model is expanded to full-scale dimension. Scaling effects are analysed. The model is used for safety assessment of the package to be approved.
Transport packages for spent nuclear fuel have to be 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 BAM for the authority assessment of spent fuel behavior within the package design approval procedure of German package designs. In particular, cracks or failures in the fuel rod cladding can occur under regulatory transport conditions. These defects can cause the release of gas, volatiles, fuel particles or fragments into the package cavity and have to be considered properly in the safety analysis. Another issue is the transport of defective fuel rods. One concept is to use special canisters which can be handled like fuel assemblies. This concept requires additional assessment concerning drying, sealing and the mechanical and thermal design of such canisters. 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 simplified approaches to consider conservatively spent fuel behavior currently accepted by BAM are presented here.
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 the sealing system of transport casks for spent fuel and high radioactive waste.
Applicants are requested by BAM to provide test programmes for verification of design leakage rates of the sealing system under the possible loads.
Two test series initiated to clarify the dependency of the standard leakage rate on dynamic lid displacement as well as the dependency of the useful elastic recovery ru of a metallic seal on temperature and time, are outlined in this paper as examples of present investigations:
Sliding tests simulating a lid displacement possible by a horizontal drop of the cask are carried out at the controlled drop test facility of BAM with test flange pairs equipped with metallic Helicoflex seals.
For specification of covering values for the useful elastic recovery ru GNS has started a very comprehensive test programme with overall 70 metallic seals installed in test flanges which are stored for a period of one year at three different temperatures up to 150°C.
This paper gives an overview about the current approach of BAM in the assessment of cask tightness and informs about the status of the running test series.
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.
Type B(U) packages for the transport of radioactive material have to withstand accident conditions of transport defined in the regulations of the International Atomic Energy Agency in form of different mechanical (drop) tests with a subsequent thermal test. According to the regulatory requirements the orientation of the package in drop tests shall be such to cause the most damaged state in the components performing the safety functions.
For the package Iid system a 9 m drop onto the unyielding target with Iid side downwards is often the most damaging orientation. The impact Ioads acting on the Iid in this orientation result mainly from interaction between Iid and internal content. In case of a movable content its impact onto the inner side of the Iid can cause additional Ioad peaks on the Iid and the Iid bolts. The intensity of the internal collision depends on the position of content relating to Iid at the time of package first contact with target. Due to physical limitations an axial gap, which could be set in "pre-drop" configuration of package or which could spontaneously appear during the drop test, usually does not cover the maximum size possible in specific package design. In this context, the combination of drop tests with post-test analysis can be helpful to better estimate the effect of internal impact. The paper summarizes some aspects of this issue based on the BAM experience in the design assessment of Type B(U) transport packages. Additionally the paper shall support applicants in German approval procedures to reduce rounds of questions and ensure delivery of reliable safety case documents to the authorities.
Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the regulations of the International Atomic Energy Agency. 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 capabilities under these conditions. Cracks or failures in the fuel rod cladding can cause the release of gas, volatiles or fuel particles into the cavity. The amount of substances in the cavity has to be considered in the assessment of the activity release and criticality safety.
The mechanical analysis of the compound system formed by the fuel rod cladding and the spent fuel pellets is very difficult due to the limited knowledge of the material properties and the insufficient understanding of the interaction between pellets and cladding and between adjacent pellets. The variation of fuel assembly properties regarding cladding material, burn-up and the history of usage makes reliable predictions of the fuel rod behavior even harder.
For a better understanding about the behavior of spent fuel rods, JRC-ITU and BAM have started a joint research project. In this context, JRC-ITU has developed a test device which allows quasi-static 3-point-bending test on fuel rod segments in the hot cell. The loads are applied with respect to the boundary conditions of the activity release assessment. This paper deals with the numerical calculation of a single fuel rod segment under bending load. The aim is to identify the governing mechanical parameters by the variation of constitutive assumptions, contact conditions, inner constraints, etc. This knowledge helps for the interpretation of the experimental results. Furthermore, the improved understanding about the behavior of the cladding-pellets system will be beneficial for the assessment of spent fuel transport conditions.
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.
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.
Trunnion systems of packages for the transport of radioactive materials have to guarantee the safe handling of package during crane operations (lifting, tilting) and to secure package tie down to the transport vehicle, if the trunnions are used as attachment points during transport. The design of trunnions is based on IAEA Regulations SSR-6, the supporting Advisory Material SSG-26 and further appropriate technical standards and/or guidelines. For package approval procedures in Germany the guideline BAM-GGR 012 has to be applied. If a package is handled inside a german nuclear power plant, the standard KTA 3905 has to be fulfilled additionally.
In this paper the requirements of KTA 3905 concerning the trunnion systems as load attaching points (LAP) are discussed in connection with the recommendations in the guideline BAM-GGR 012. This guideline is prepared at BAM Federal Institute for Materials Research and Testing for analysis and assessment of bolted lid and trunnion systems of Type B(U) transport packages. The Quality assurance questions concerning trunnion systems are discussed as well.
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.
Acceptable limits for activity release from transport casks for high-level radioactive material specified in the IAEA regulations must be kept by the integrity of cask body and the cask sealing system. BAM as the German competent authority for mechanical, thermal and containment assessment of packages liable for approval verifies the activity release compliance with the regulatory limits. One of the fundamental aspects in assessment is the specification of conservative package design leakage rates.
To ensure the required package tightness for both, storage, and transport of the cask before and after storage usually metal seals of the Helicoflex® Type are used. Due to the long-term use the seal behavior is influenced by temperature and time. The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport specified in the regulations can have a significant effect on the leak tightness of the sealing system. Whereas the safety for application of new, non- aged Helicoflex® seals is verified sufficiently, there are still technical data gaps concerning the efficiency of aged Helicoflex® seals.
BAM performed experiments to learn more about the sealing efficiency of aged Helicoflex® seals with Aluminum and Silver outer jackets. The seals were compressed in test-flanges and for artificial ageing the complete flange systems were stored in an oven for several month at a high temperature. During the compression and decompression tests after the aging, load-deformation characteristics of the seals, and leakage rates were measured. With these tests a load situation was simulated, which can occur in the regulatory drop test of the cask: Under high impact loads the bolted lid can lift a little for a short moment, allowing a little movement of the seal, so that the contact area can change before compressing again.
The poster presentation will show details about test conditions and first results.
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. Generally, the mechanical behavior of high burn-up spent fuel assemblies under transport conditions shall be analyzed with regard to the assumptions which are used in the containment and criticality safety analysis. Considering 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. 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 additionally. As a result, enveloping analytical approaches are usually used by BAM within the safety assessment of packages approved for transport of spent nuclear fuel.
To justify the safety margins of such approaches additional analyses are necessary. In this paper, numerical simulations of a spent fuel assembly Segment are presented. The segment modeled represents the part of a generalized BWR fuel assembly between two spacers.
Dynamic and quasi-static finite element calculations are performed to simulate the spent fuel behavior under regulatory defined accident conditions of transport. Beam elements are used for the modeling of the fuel rods representing the compound consisting of claddings and fuel pellets. The dynamic load applied is gathered from an experimental drop test with a spent fuel cask performed at BAM. A hot cell bending test performed at JRC Karlsruhe is the basis for obtaining the material behavior of the fuel rods. The material properties are determined by simulating the test setup of JRC and optimizing the results to fit the experimental load deflection curve. The simulations of the fuel assembly segment are used to get a better understanding about the loads on fuel rods under accident conditions of transport.
The Federal Institute for Materials Research and Testing (BAM) is assessing the mechanical and thermal safety performance of packages for the transport of radioactive materials. Drop testing and numerical calculations are usually part of the safety case concepts, where BAM is performing the regulatory tests at their own test facility site.
Among other mechanical tests the 1 meter drop onto a steel puncture bar shall be considered for accident safe packages. According to 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…”.
Particularly with regard to the German transport- and storage cask designs, often made from ductile cast iron, an accurate determination of the puncture bar length to guarantee a load impact covering the worst case scenario can be imperative. If the fracture mechanical proof for the cask material shall be provided by a test, small deviations in the concentrated load applied can be decisive for the question if the cask fails or not.
The most damaging puncture bar length can be estimated by iterative procedure in numerical simulations. On the one hand, a sufficient puncture bar length shall 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.
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 the containment system of transport casks for spent fuel and high radioactive waste. The containment systems of such transport casks usually include bolted lids with metallic or elastomeric seals.
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 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 and open points of the transferability of scaled test results to the fullscale design of the containment system will be discussed.
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.
For transport package design and operation according to the IAEA regulations, the package shall be securely stowed and its retention system shall be capable to withstand load conditions of routine transport. The supporting IAEA Advisory Material SSG-26 provides information how to do that.
Up to now package designers in different countries use other load factors for the design of attachment points than those specified in the IAEA guidance material. In particular the acceleration values vary between different countries and lead to difficulties during the validation of foreign approval certificates. Therefore the IAEA started a discussion process to review the existing guidance text. An international working group was constituted in 2013. Representatives came from different stakeholders, e.g. transport operators, competent authorities and modal organizations. The discussions concluded especially on the transport conditions which has to be considered for stowage design, including on the one hand the relevance of the load factors used for strength and fatigue analysis and on the other hand the criteria which have to be considered for the attachment points.
The proposed acceleration values will be compared to those measured during recent multi-modal testing by Sandia National Laboratories that measured the acceleration levels experienced by a spent fuel flask during heavy-haul truck, sea, and rail transport.
The ISO standard 10276 is dealing with the load attachment systems of packages as well. This standard considers the trunnion design, manufacturing and operational aspects. The regular standard revision phase started in 2017. An expert group discussed new state-of-the-art technology, different analysis approaches for strength and fatigue analysis and proposed revised text for the ISO standard for international discussion. The finite-element analysis approach incl. appropriate acceptance criteria are described and referenced.
The paper describes relevant tie-down aspects, gives background argumentation relevant to analysis approaches, and tries to support harmonized application of the revised IAEA guidance material and the future revised ISO standard.
The containment systems of transport and storage casks for spent fuel and highlevelradioactive 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.
For disposal of the German research reactor of the Technical University Munich FRM II a new transport and storage cask design was under approval assessment by the German authorities. The Bundesanstalt für Materialforschung und -prüfung (BAM) assessed the mechanical and thermal package safety and performed drop tests. The activity release approaches and subjects of quality assurance and surveillance for manufacturing and operation of the package were assessed by BAM as well.
The cask body is made of ductile cast iron and closed by two bolted lid systems with metal seals. The material of the lids is stainless steel. On each end of the cask a wood-filled impact limiter is installed to reduce impact loads to the cask under drop test conditions. In the cavity of the cask a basket for five spent fuel assemblies is arranged.
For the safety case a combination of experimental testing and analytical/numerical calculations were applied. In total, four drop tests were carried out at the BAM large drop test facility. Two tests were carried out as a full IAEA drop test sequence consisting of a 9m drop test onto an unyielding target and a 1m puncture bar drop test. The other two drop test were performed as single 9m drop tests and completed by additional analyses for considering the effects of an IAEA drop test sequence.
The main objectives of the drop tests were the investigation of the integrity of the package and its safety against release of radioactive material as well as the test of the fastening system of the impact limiters. Furthermore, the acceleration and strain signals measured during the tests were used for the verification of Finite-Element-Analysis (FEA) used for the safety analysis of the package design.
The finite-element models incorporated in the package design safety report include the cask body, the lid system, the inventory and the impact limiters with the fastening system. In this context special attention was paid to the modeling of the encapsulated wood-filled impact limiters.
Additional calculations using the verified numerical models were done by the applicant and assessed by BAM to investigate e.g. the brittle fracture of the cask body made of ductile cask iron within the package design approval procedure.
This paper describes the package design assessment from the view of the competent authority BAM including the applied assessment strategy, the conducted drop tests and the additional calculations by using numerical and analytical methods.