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Organisationseinheit der BAM
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.
BAM Federal Institute for Materials Research and Testing is the competent authority for mechanical safety assessment of transport packages for radioactive material in Germany. The further development of state-of-the-art technology concerning assessment methods is essential for a qualified work of involved designers and authority experts. The paper gives an example of current development done to improve understanding and modeling capabilities of wood filled impact limiter. In order to reduce the loads applied to the package containment, which result from regulatory drop tests, most packages are protected by energy dissipating impact limiter. Wood, encapsulated by steel sheets, is one of the materials typically used for energy dissipation in these impact limiter. Very often, mechanical safety cases regarding the 9 m drop test are performed computationally, where it is essential to use reliable and verified computational methods and models. In this context, the paper presents an approach for a finite element material model for wood. Thereby, the mechanical behavior of wood under compression loading is the focus of the development work. Additionally, material orientation as well as strain rate, temperature and lateral constraint may vary. A large number of experiments, particularly compression tests, was designed and performed to establish an adequate experimental database for modeling verification. The experimental results enabled the derivation of necessary requirements: The material model has to take into account strain rate and temperature dependencies as well as the anisotropic characteristics of the material, a proper yield criterion, flow rule and hardening law. Such a material model is currently not available in established commercial dynamic finite element codes. Thus it is necessary to create a user-defined material model considering the mentioned requirements. A first step was done by determining a yield surface as well as detecting flow and hardening mechanisms from experimental force-deflection curves. In a next step the LS-DYNA material model MAT_75 was altered according to conclusions of former BAM development work, regarding the modeling of post-peak softening as a function of lateral constraint. Future research will contain the further development, implementation and verification of a material model for wood.
The crush test for light weight and low density type B packages was introduced for the first time into the 1985 edition of the International Atomic Energy Agency (IAEA) transport safety regulations. In the early 1970s, the need for an additional mechanical test besides or instead of the well known 9 m drop test was deliberated. Various authors and test facilities, including BAM and Sandia National Laboratories (SNL), were able to prove that the level of safety provided by IAEA drop and puncture tests in the regulations did not protect against dynamic crush forces to smaller packages. As early as the third PATRAM symposium held in 1971 (Richland, WA, USA), Robert F. Barker asked for '... a more strenuous crushing test for protecting small, light weight packages ...' BAM developed from research activities a proposal as to which types of packages should be subject to crush tests and how the crush tests should be performed, which was presented at the 5th PATRAM symposium held in 1978 (Las Vegas, NV, USA). At the IAEA, the possible need for a crush test was first mentioned in 1977. The subject for a discussion, besides the principal need for this test, was also the development of suitable set of crush test boundary conditions. It took more than four years of discussion until a dynamic crush test similar to today's test was recommended by experts to the IAEA regulatory revision panel. Finally, after a rigorous evaluation process in which also the boundary conditions were determined, the crush test was proposed to be incorporated into the IAEA regulations. BAM and SNL participated in the crush test development and implementation process right from the beginning in the early 1970s until its implementation in the IAEA regulations in 1985. Today, BAM performs crush test procedures according to para. 727(c) of TS-R-1, which have not been changed since their first implementation. Crush tests performed in 2002 at BAM will be discussed. These approval design tests were performed on birdcage pellet transport containers under normal and accident conditions according to the IAEA regulations.
In the approval procedure of transport packages for radioactive materials, the mechanical and thermal safety assessment is carried out in Germany by competent authority BAM. In recent years BAM was involved in several licensing procedures of new spent fuel and HLW package designs, where the cask body was made of Ductile Cast Iron (DCI). According to IAEA regulations package designs have to fulfill requirements for specific conditions of transport. Type B(U) packages must withstand the defined accident conditions of transport. The temperature range from -40°C up to the operational temperature has to be considered. For the cask material DCI, it is necessary to determine the brittle fracture behavior. The German guideline BAM-GGR 007 defines requirements for the fracture mechanics of DCI. Due to complex structure of the cask body and the dynamic loading a fracture mechanical assessment in an analytical kind is not always possible. Experience of recent design approval procedures show that the application of numerical calculations are applicable to determine the fracture mechanical load in the cask body. At the first step a numerical analysis has to be done to identify the loading state at the whole cask body. Secondly an analysis of a detail of the cask body is made considering the displacement boundary conditions of the global model. An artificial flaw is considered in this detailed model to calculate the fracture mechanical loading state. The finite element mesh was strongly refined in the area of flaw. The size of the artificial flaw is characterized by the ultrasonic inspection used for the quality assurance of the package. The applicant developed additional analysis tools for calculation of stress intensity factor and/or J-Integral. The assessment approach by BAM led to the decision to develop own tools to the possibility for independent check of the results.
The paper describes the authority assessment approach for the DCI fracture mechanics analysis. The validation procedure incl. the development of own tools is explained. BAM developed a post-processor called JINFEM to determine the fracture mechanical loads. The regulatory 1 m puncture bar drop test is used to give an example of the assessment procedure.
The moisture content of wood is known to have a significant influence on the woods mechanical properties. Using wood as an energy absorber in impact limiter of packages for the transport of radioactive material, it is of particu-lar importance to ensure the moisture content and thus relevant mechanical properties to be in specified limits. The paper surveys the influence of wood moisture content on the mechanical properties of wood. Different measuring methods are discussed with respect to in-situ applicability, accuracy and effort. The results of an experimental ana-lysis of the accuracy of hand-held moisture meters using the electrical resistance method are discussed. Conclu-sions are drawn regarding the measurement of moisture content of wood upon delivery as well as of complete im-pact limiter assemblies. Requirements for quality surveillance during manufacturing of wood filled impact limiter are derived and it is exemplified how to meet them. Construction, manufacturing and inspection of impact limiter encapsulation with regard to leak-tightness are addressed.
Using Scale Model Impact Limiter in the Type Assessment of Transport Casks for Radioactive Material
(2007)
Simulation of the Crushing of Wood Filled Impact Limiters for Packages of Radioactive Material
(2008)
Mechanical and thermal safety assessment of packages for transport of radioactive material in Germany is carried out by the Federal Institute for Materials Research and Testing (BAM). Both experimental and computational (analytical, numerical) methods combined with material and/or component tests are the basis for the state of the art safety assessment concept at BAM. The required mechanical tests according to IAEA regulations include, among others, a 9-m-drop-test on an unyielding target. Impact limiting components, which are attached to the cask at both ends, limit forces applied on the cask body and lid system by absorbing a major part of the impact energy. In Germany, impact limiters of packages for transport of radioactive materials are typically of steel-wood-sandwich construction, combining a relatively stiff steel structure bolted to the cask body, outer steel plates and different types of wood. By crushing the wood-steel-sandwich-structure between an unyielding target and the cask, the kinetic energy of a 9-m-free-fall is absorbed. The main energy absorber is wood under a high level of deformation. Wood under large deformations exhibits destruction of the fibre matrix. By analysing compression of the impact limiter wood after the drop tests with prototype casks for radioactive material, underlying mechanisms of wood crushing and corresponding energy absorption under large deformations are identified. Softening occuring at compression of the wood is a function of the lateral strain restriction of wood. Against the background of continuum mechanics an analogous model for compression of the fibre bundle is presented. The model takes the lateral strain restriction as triaxiality of the stress state into account. Further modelling possibilities for wood with a continuum approach are described. Different material laws in the explicit finite element code LS-DYNA are analysed for possible application using the analogous model for the fibre bundle. Small scale compression tests with wooden specimens are modelled in order to evaluate the ability of different modelling techniques to simulate softening. Although modelling of the compression of wood under large deformations is possible, softening could not be simulated purposefully. A drop test of a cask with impact limiting devices similar to existing impact limiters is simulated with different material laws for wood. The behaviour of impact limiting devices could not be simulated universally including the influence of the lateral strain restriction; nevertheless loading of the cask by crushing of the impact limiter could be simulated purposefully. Verification with experimental results is essential.
The moisture content of wood is known to have a significant influence on the wood's mechanical properties. Using wood as an energy absorber in impact limiters of packages for the transport of radioactive material, it is of particular importance to ensure the moisture content and thus relevant mechanical properties to be in specified limits. The paper surveys the influence of wood moisture content on the mechanical properties of wood. Different measuring methods are discussed with respect to in situ applicability, accuracy and effort. The results of an experimental analysis of the accuracy of hand held moisture metres using the electrical resistance method are discussed. Conclusions are drawn regarding the measurement of moisture content of wood upon delivery as well as of complete impact limiter assemblies. Requirements for quality surveillance during manufacturing of wood filled impact limiters are derived and it is exemplified how to meet them. Construction, manufacturing and inspection of impact limiter encapsulation with regard to leak tightness are addressed.
Admissible limits for activity release from type B(U) packages for spent fuel transport specified in the International Atomic Energy Agency regulations (10-6 A2 h-1 for normal conditions of transport and A2 per week for accidental conditions of transport) have to be kept by an appropriate function of the cask body and its sealing system. Direct measurements of activity release from the transport casks are not feasible. Therefore, the most common method for the specification of leak tightness is to relate the admissible limits of activity release to equivalent standardised leakage rates. Applicable procedure and calculation methods are summarised in the International Standard ISO 12807 and the US standard ANSI N14·5. 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. Two fundamental aspects in the assessment are the specification of conservative design leakage rates for normal and accidental conditions of transport and the determination of release fractions of radioactive gases, volatiles and particles from spent fuel rods. Design leakage rates identify the efficiency limits of the sealing system under normal and accidental transport conditions and are deduced from tests with real casks, cask models or components. The releasable radioactive content is primarily determined by the fraction of rods developing cladding breaches and the release fractions of radionuclides due to cladding breaches. The influence of higher burn-ups on the failure probability of the rods and on the release fractions are important questions. This paper gives an overview about methodology of activity release calculation and correlated boundary conditions for assessment.
Impact limiters of packages for transport of radioactive materials are in Germany typically steel-wood-sandwich-constructions, combining a relatively stiff steel structure bolted to the cask body, outer steel plates and different types of wood. By crushing the wood-steel-sandwich-structure between unyielding target and cask, kinetic energy of 9-m-free-fall is absorbed. Main energy absorber is wood under a high level of deformation. Wood under large deformations exhibits destruction of the fibre matrix. By analysing compression of impact limiter wood after drop tests with prototype casks for radioactive material, underlying mechanisms of wood crushing and corresponding energy absorption under large deformations are identified. Softening occurring at compression of wood is a function of lateral strain restriction of wood. Against the background of continuum mechanics an analogous model for compression of fibre bundle is presented. The model takes lateral strain restriction as triaxiality of stress state into account. Further modelling possibilities for wood with a continuum approach are described. Different material laws in the explicit Finite Element code LS-DYNA are analysed for possible application on the analogous model for the fibre bundle. Small scale compression tests with wooden specimens were modelled in order to evaluate the ability of different modelling techniques to simulate softening. Although modelling of compression of wood under large deformations is possible, softening could not be simulated purposefully. A drop test of a cask with impact limiting devices similar to existing impact limiters is simulated with different material laws for wood. Behaviour of impact limiting devices could not be simulated universally including influence of lateral strain restriction; nevertheless loading of the cask by crushing of impact limiter could be simulated purposefully. Verification with experimental results is compulsory.
Transport of large nuclear power plant components: experiences in mechanical design assessment
(2009)
In the course of decommissioning of power plants in Germany large nuclear components (steam generator, reactor pressure vessel) must be transported over public traffic routes to interim storage facilities, where they are dismantled or stored temporarily. Since it concerns surface contaminated objects or low specific activity materials, a safety evaluation considering the IAEA transport regulations mainly for industrial packages (type IP-2) is necessary. For these types of industrial packages the requirements from normal transport conditions are to be covered for the mechanical proof. For example, a free drop of the package from a defined height, in dependence of its mass, onto an unyielding target, and a stacking test are required. Since physical drop tests are impossible generally due to the singularity of such 'packages', a calculation has to be performed, preferably by a complex numerical analysis. The assessment of the loads takes place on the basis of local stress distributions, also with consideration of radiation induced brittleness of the material and with consideration of recent scientific investigation results. Large nuclear components have typically been transported in an unpackaged manner, so that the external shell of the component provides the packaging wall. The investigation must consider the entire component including all penetration areas such as manholes or nozzles. According to the present IAEA regulations the drop position is to be examined, which causes the maximum damage to the package. In the case of a transport under special arrangement a drop only in an attitude representing the usual handling position (administratively controlled) is necessary. If dose rate values of the package are higher than maximum allowable values for a public transport, then it is necessary that additional shielding construction units are attached to the large component.
Mechanical Design Assessment Approaches of Actual Spent Fuel and HLW Transport Package Designs
(2010)
Basierend auf Fallversuchen mit Transportbehältern für radioaktive Stoffe sowie auf einem Versuchsprogramm mit Holzproben wurden die Energieabsorptionsmechanismen identifiziert und ein Modell für das Verhalten von Fichtenholz bei axialer Belastung entwickelt. Das Modell zieht für die Kompression von Holz die seitliche Dehnungsbehinderung – im kontinuumsmechanischen Zusammenhang wäre das die Mehrachsigkeit des Spannungszustandes im Kontinuum – in Betracht. Die Energie, die vom Holz absorbiert werden kann, ist umso größer, je größer die seitliche Dehnungsbehinderung ist.
Für die Modellierung mit Finite Elemente Methoden wurden verschiedene Modellierungsstrategien untersucht, keine der vorgeschlagenen Modellierungsstrategien war in der Lage, dass Verhalten von Holz bei Druckbeanspruchung und großen Deformationen inkl. der Entfestigung zu modellieren. Bei Verwendung einer entfestigenden Fließkurve zeigt das Modell ausgeprägte Netzabhängigkeiten und numerischen Instabilitäten.
Für die Modellierung der Entfestigung mit Hilfe einer von der Mehrachsigkeit des Spannungszustandes gesteuerten Fließflächenevolution ist kein geeignetes Materialmodell vorhanden.
Anhand der Simulation eines Fallversuchs mit einem Behälterkörper und stoßdämpfenden Bauteilen im Maßstab 1:2 wurde gezeigt, dass zwar eine vollständige Modellierung des Stoßdämpferverhaltens inkl. Entfestigung nicht möglich war, aber das Stoßdämpferverhalten mit einer nicht-entfestigenden Fließkurve sinnvoll modelliert werden konnte. Eine Verifikation des Modellierungsansatzes am real zu modellierenden Bauteil ist zwingend erforderlich. Wenn signifikante Änderungen zwischen Verifikations- und Simulationsobjekt auftreten, ist die Zuverlässigkeit der Rechenergebnis als gering zu bewerten. Nur wenn die auftretenden Kompressionsmechanismen durch eine Analyse des Stoßdämpferholzes bestimmt werden können, ist eine zuverlässige Ermittlung von Stoßdämpferkräften- und verformungen möglich.
Transport casks for radioactive materials have to withstand the 9 m drop test, 1 m puncture drop test and dynamic crush test with regard to the mechanical requirements according to the IAEA regulations. The safety assessment of the package can be carried out on the basis of experimental investigations with prototypes or models of appropriate scale, calculations, by reference to previous satisfactory safety demonstrations of a sufficiently similar nature or a combination of these methods. Computational methods are increasingly used for the assessment of mechanical test scenarios. However, it must be guaranteed that the calculation methods provide reliable results. Important quality assurance measures at BAM are given concerning the preparation, run and evaluation of a numerical analysis with reference to the appropriate guidelines.
Hence, a successful application of the finite element method requires a suitable mesh. An analysis of the 1 m puncture drop test using successively refined finite element meshes was performed to find an acceptable mesh size and to study the mesh convergence using explicit dynamic finite element codes. The finite element model of the cask structure and the puncture bar is described. At the beginning a coarse mesh was created. Then this mesh was refined in two steps. In each step the size of the elements was bisected. The deformation of the mesh and the stresses were evaluated dependent on the mesh size. Finally, the results were extrapolated to an infinite fine mesh or the continuous body, respectively. The uncertainty of the numerical solution due to the discretization of the continuous problem is given. A safety factor is discussed to account for the uncertainty. The calculation results are compared with experimental data from a puncture drop test with a half-scale model of a cylindrical cask.
This paper supports the convergence studies of the Task Group on Computational Modeling for Explicit Dynamics reporting to the ASME BPV Code Working Group on Design Methodology.
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.
In Germany, the mechanical and thermal safety assessment of approved packages for the transport of RAM is carried out by BAM as the competent authority according to the International Atomic Energy Agency regulations. BAM was involved in several approval procedures with ductile cast iron containers containing wet intermediate level waste. These contents, which are not dried, only drained, consist of saturated ion exchange resin and a small amount of free water. Compared to the safety assessment of packages with dry content, attention must be paid to some more specific points. The physical and chemical compatibility of the content itself and of the content with materials of the package must be shown. From the mechanical resistance point of view, the package has to withstand the forces resulting from the freezing liquid. The most interesting point, however, is the pressure build-up inside the package due to vapourisation. This could be caused by radiolysis of the liquid and must be taken into account for the storage period. The paper deals primarily with the pressure build-up inside the package caused by the regulatory thermal test (30 min at 800°C) as part of the cumulative test scenario under accident conditions of transport. To determine the pressure, the temperature distribution in the content must be calculated for the whole period from the beginning of the thermal test until cooling down. In this case, calculating the temperature distribution requires, besides the consideration of conduction and heat radiation, consideration of evaporation and condensation including the associated processes of transport.
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.
Thermal test driven pressure build-up inside type-B packages containing wet radioactive waste
(2013)
In recent years several German approval procedures for ductile cast iron transport containers containing wet intermediate level waste were conducted. BAM, as one of the German competent authorities, was involved in the complex design assessment work with this specific issue. Thermal analysis is one part of the authority assessment work done by BAM in Germany. The radioactive contents of package designs which were not dried, only drained, consist of saturated ion exchange resin and a small amount of free water. Compared to the safety assessment of packages with dry content, attention must be paid to some more specific points. The most interesting point, however, is the pressure build-up inside the package due to vaporization. This could be caused by radiolysis of the liquid and must be taken into account for the storage period. The inner pressure of the package leads to mechanical loads to the package body, the lid and the lid bolts. Thus, the pressure is the driving force on the gasket system regarding the activity release and a possible loss of tightness. The paper deals primarily with the pressure build-up inside the package caused by the transport regulatory thermal test (30 min at 800 °C) as part of the cumulative test scenario under accident conditions of transport. The pressure build-up is estimated by calculation in a very conservative way regarding conduction and heat radiation. Furthermore the paper discusses a conservative approach for the estimation of the resulting pressure depending on the percentage of water inside the cask. To get trustworthy results without an exact specification of the content, experimental fire tests should be conducted. However, this paper shows the difficulties of assessing casks containing wet content. From the authority assessment point of view, drying of the content could be an effective way to avoid the above described pressure build-up and the associated difficulties for the safety assessment.
In the design approval of transport packages for radioactive materials, the mechanical and thermal safety assessment is carried out in Germany by competent authority BAM. In recent years BAM was involved in several licensing procedures of new spent fuel and HLW package designs, where the cask body is of Ductile Cast Iron (DCI). According to IAEA regulations package designs have to fulfill requirements for specific conditions of transport. Type B(U) packages must withstand the defined accident conditions of transport. The temperature range from -40°C up to the operational temperature has to be considered. For the cask material DCI, it is necessary to determine safety against brittle fracture. The German guideline BAM-GGR 007 defines requirements for fracture mechanics of packagings made of DCI. Due to complex cask body structure and the dynamic loading a fracture mechanical assessment by analytical approaches is not always possible. Experience of recent design approval procedures show that the application of numerical calculations are applicable to determine the stresses and stress intensity factors in the cask body. At the first step a numerical analysis has to be done to identify
the loading state at the whole cask body. Secondly an analysis of a detail of the cask body is made considering the displacement boundary conditions of the global model. An artificial flaw is considered in this detailed model to calculate the fracture mechanical loading state. The finite element mesh was strongly refined in the area of the flaw. The size of the artificial flaw is based on the ultrasonic inspection acceptance criteria applied for cask body manufacture. The
applicant (GNS) developed additional analysis tools for calculation of stress intensity factor and/or J-Integral. The assessment approach by BAM led to the decision to develop own tools to the possibility for independent proof of the results.
The paper describes the authority assessment approach for DCI fracture mechanics analysis. The validation procedure incl. the development of own tools is explained. BAM developed a postprocessor to determine the fracture mechanical loads. A horizontal 1 m puncture bar drop test is used to give a detailed description of the assessment procedure.
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.
Für die Beförderung von umschlossenen radioaktiven Stoffen (Strahler) können Transporterleichterungen gelten, wenn das Strahlerdesign nachweislich unfallsicher ausgelegt ist und eine Zulassung als radioaktiver Stoff in besonderer Form vorliegt. Die in Deutschland zuständige Behörde für die Prüfung und Zulassung radioaktiver Stoffe in besonderer Form ist die Bundesanstalt für Materialforschung und –prüfung (BAM).
Eine oft langfristige Nutzung der Strahler kann eine alterungsbedingte Schwächung der Umschließung des radioaktiven Stoffes zur Folge haben. Jeder Strahler einer zugelassenen Bauart muss jedoch im Beförderungsfall zu jeder Zeit, auch nach längerer Nutzung, den vorgeschriebenen Prüfanforderungen genügen. Die Alterungsbewertung auf der Grundlage einer vom Antragsteller zu spezifizierenden Nutzungsdauer des radioaktiven Stoffes in besonderer Form ist seit langer Zeit Bestandteil des Zulassungsverfahrens in Deutschland. Ein von Deutschland eingebrachter Vorschlag im Rahmen der anstehenden Revision der IAEA Empfehlungen für die sichere Beförderung von radioaktiven Stoffen (SSR-6) soll die Lesbarkeit des Regelwerkes bezüglich der Alterungsbewertung verbessern, Klarheit über die damit verbundenen Anforderungen erzeugen und damit zu einer Harmonisierung der Zulassungsverfahren auf internationaler Ebene beitragen. In diesem Beitrag sollen wichtige Einflussfaktoren auf das Alterungsverhalten von radioaktiven Stoffen in besonderer Form aufgezeigt und die Notwendigkeit für die Spezifikation einer Nutzungsdauer als Grundlage für die Alterungsbewertung begründet werden. Der deutsche Vorschlag für die Revision des IAEA Regelwerks zum sicheren Transport radioaktiver Stoffe (SSR-6) wird vorgestellt und erläutert.
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.
CASTOR-Behälter sollen den Auswirkungen schwerster Unfälle standhalten und müssen dabei ihre Sicherheitsfunktionen beibehalten. Dazu zählen der sichere Einschluss des radioaktiven Inventars mit nachgewiesener Behälterintegrität und -dichtheit, die ausreichende Abschirmung der radioaktiven Strahlung, die Ableitung der von den Brennelementen ausgehenden Wärme und der Ausschluss des Entstehens einer nuklearen Kettenreaktion, d.h. die Kritikalitätssicherheit sind zu gewährleisten.
Zum Nachweis dieser Fähigkeiten werden die Behälter gegen Stoß, Aufprall, Durchstoßen, Feuer sowie beim Eintauchen in Wasser geprüft. Durch diese Tests werden Beanspruchungen aus potentiellen schweren Unfällen abdeckend simuliert. Die hierfür erforderlichen experimentellen Tests, aber auch die Prüfung und Entwicklung komplexer rechnerischer Simulationen des Behälterverhaltens unter Anwendung der Finite-Elemente-Methode erfolgen bei der Bundesanstalt für Materialforschung und -prüfung (BAM).
Die Basis bildet das Regelwerk der Internationalen Atom- und Energieorganisation (IAEO) zum sicheren Transport von radioaktiven Stoffen, welches international harmonisierte Schutzziele und Prüfanforderungen definiert. Die seit 60 Jahren praktizierte stetige Weiterentwicklung der Sicherheitsanforderungen trägt aktuellen Erkenntnissen, Erfahrungen und Risikoanalysen Rechnung. Die Brennelementbehälter werden mechanischen Prüfungen sowie Brand- und Wasserdruckprüfungen unterzogen. Hierdurch wird nachgewiesen, dass sie den Auswirkungen eines schweren Unfalls standhalten können.
Nur Behälter die zweifelsfrei die kumulativen mechanischen und thermischen Tests und auch die Wasserdruckprüfung mit Erfüllen aller Schutzziele bestehen, bekommen eine Zulassung, die es erlaubt die Behälter zu transportieren.
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.
With integration of the new para 613A into SSR-6 [1] the consideration of ageing mechanisms is now obligatory for the design of packages and their approval. In addition, para 809(f) of SSR-6 [1] requires consideration of the effects of storage on ageing mechanisms, safety analyses and operation and maintenance instructions.
German competent authorities Bundesanstalt für Materialforschung und -prüfung (BAM) and Bundesamt für die Sicherheit der nuklearen Entsorgung (BASE) are considering the aspect of ageing in approval procedures. Ageing assessment is mainly focused on dual purpose casks (DPC) package designs which are long-term stored in interim storage facilities. For these package designs, the evaluation of ageing management is now mandatory for the maintenance of the package design approvals with a validity period of 5 years and beyond. The ageing management includes amongst others a gap analysis, the assessment of ageing effects and operational experiences during operation and interim storage.
BAM works on the compilation of a guideline for implementation of paras 613A, 809(f) and 809(k) for packages requiring competent authority approval at the application procedure in Germany. The paper describes essential items of ageing mechanisms and will give a foresight to the ageing management evaluation by BAM.
Transport packages shall satisfy various safety criteria regarding mechanical, thermal and radiation phenomena. Typical requirements focusing mechanical aspects are usually drop tests in accordance with IAEA regulations. The drop tests are usually carried out experimentally and, as an additional measure, finite element analyses (FEA) are performed. A specific part of the investigations presented is the evaluation of the welding seam connecting cask shell and cask bottom. Experimental results and FEA are presented and compared. The evaluation of the welding seam performed includes a variety of aspects. In addition to the experimental and analytical stresses determined, different standards are used to investigate a possible crack initiation. Several destructive and non-destructive tests are performed for quality assurance in the manufacturing process as well as for different input parameters. The necessary monitoring and non-destructive measurement methods to define the boundary conditions of the standards are introduced. Taking into account all required parameters, the welding seam is examined and evaluated using the failure assessment diagrams (FAD) of the respective standards. It can be shown under the given boundary conditions that considering the experimental data, the welding seam is in the context of crack initiation below the enveloping curve in the acceptable region. More critical drop tests to be conducted are proposed and need to be investigated in future work.
The decommissioning or refurbishment of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, and coolant pumps, to list the major contributors. These components or objects are large in size and mass, measuring up to approximately 6 meters in diameter, up to 20 meters in length, and weighing over 400 000 kg. In many situations, the components are transported off-site to a storage, disposal or recycling/treatment facility. Previously, many large objects had to be transported under special arrangement.
The latest 2018 edition of the International Atomic Energy Agency (IAEA) Regulations for the Safe Transport of Radioactive Material, No. SSR-6 [1], incorporates regulations for the shipment of large objects as a new category of surface contaminated object, SCO-III, based on the IAEA “performance package” concept. This paper provides background and practical guidance on these regulations.
Additionally, the experiences of BAM with the appoval of two steam converters of the NPP Lingen are presented as the first approval process for SCO-III objects in Germany.
The primary additions to SSR-6 include SCO-III classification and requirements, approval and administrative requirements for the new classification, and the addition of SCO-III to the proper shipping
name for UN 2913. Advisory material drafted for the new requirements will be included in the next revision of SSG-26, Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, expected to be published soon.
Note that at this time the proposed provisions for large objects do not include components such as reactor vessels, due to the more limited experience and greater radioactivity levels. The SCO-III concept lays the groundwork and may be extended to cover other large objects that are classified as low specific activity (LSA) material in the future.
Im Zuge des Rückbaus kerntechnischer Anlagen in Deutschland müssen u. a. Großkomponenten (Dampferzeuger, Reaktordruckbehälter) über öffentliche Verkehrswege in Zwischenlager transportiert werden, in denen sie zerlegt bzw. zwischengelagert werden. Da es sich hierbei um aktivierte und/oder oberflächenkontaminierte Objekte handelt, ist eine Begutachtung unter Berücksichtigung der Gefahrgutbeförderungsvorschriften notwendig. In den meisten Fällen handelt es sich um oberflächenkontaminierte Gegenstände, welche als Industrieversandstücke des Typs IP-2 zu befördern sind.
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.
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.
Dicht und sicher abgeschirmt
(2014)
Existing spent nuclear fuel (SF) and high active waste (HAW) management policies and practices worldwide are the result of past presumptions that sufficient reprocessing and/or disposal capacity would be available in the near term. Consequently, in the past many countries have developed specific solutions for different periods of time due to their individual national nuclear policies.
In Germany the concept of dry interim storage in dual purpose metal casks before disposal is being pursued for SF and HAW management and transport and storage licenses have been issued accordingly. The current operation licenses for existing storage facilities have been granted for a storage period of up to 40 years. This concept has demonstrated its suitability for over 20 years so far. Relevant safety requirements haven been assessed for the short-term as well as for the long-term for site-specific operational and accidental storage conditions. But in the meantime significant delays in the national repository siting procedure occured which will make extended storage periods necessary in the future.
This paper describes the current situation in Germany with regard to dry cask storage and focuses on current perspectives considering regulatory, technical, and scientific aspects for storage license renewal. Since there is one case of a storage license limitation to only 20 years due to administrative reasons, first experience is currently gathered in case of an expiring storage license. Subsequent license options have been pursued intensively including the extension of the initial storage license as well as shipping all casks to another storage facility considering an extended storage period at that storage facility as well. All safety relevant aspects have to be reviewed on basis of the current state-of-the-art which might be different from the initial safety demonstrations. That includes new safety assessment standards as well as improved knowledge base. Major issues are e.g. improved accident scenario analyses, assessment methods, and consideration of aging effects from previous operation periods. Inspection programs with respect, e.g. to bolts and seals to verify leak-tightness and lid screw pre-stress have been initiated to demonstrate proper cask conditions for extended storage as well as transportation to another storage facility.
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.
Packages for the transport of SNF and HLW are usually equipped with impact limiters to reduce the loads that result from the regulatory 9 m drop test. A common impact limiter design in Germany is a welded steel sheet structure filled with wood. The material wood is the main energy absorber, while the steel sheet provides the integrity of the impact limiter. The IAEA allows mechanical safety cases of transport packages to be carried out computationally, as long as the models used are reliable. In this context, a Finite Element (FE) modeling approach for wood and its application to impact limiters in the calculation of a 9 m drop test is presented.
A user material model for wood was developed for the dynamic FE-Code LS-DYNA. Its features are based on a series of crush tests with spruce wood specimens. The model considers wood as a material with transversely isotropic properties, i.e. in the directions parallel and perpendicular to the fiber. The plastic material behavior depends on the state of stress. This has shown to be important to account for the lateral constraint of wood in impact limiters resulting from steel sheet encapsulation. Lateral constraint or respectively, a multiaxial stress state, increases the compression strength level of wood, limits the softening effect and increases the hardening effect. Lateral constraint also increases volumetric and reduces deviatoric deformation. The wood material model considers various hardening and softening characteristics via input flow curves. It considers effects of temperature and strain rate on strength as well. The development of a multi-surface yield criterion and a plastic potential that enables the user input of plastic Poisson's ratios were the challenges during the development of material model.
A dynamic FE calculation of a horizontal drop test with an 18,000 kg test package was performed. The wood material model was used to model the wooden impact limiter inlays. The impact limiter deformation and the package deceleration were compared to the experimental drop test results to rate the performance of the wood material model.
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.
Mechanical design assessment approaches of actual spent fuel and HLW transport package designs
(2010)
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 casks for radioactive materials have to withstand the 9 m drop test, 1 m puncture drop test and dynamic crush test with regard to the mechanical requirements according to the IAEA regulations. The safety assessment of the package can be carried out on the basis of experimental investigations with prototypes or models of appropriate scale, calculations, by reference to previous satisfactory safety demonstrations of a sufficiently similar nature or a combination of these methods. Computational methods are increasingly used for the assessment of mechanical test scenarios. However, it must be guaranteed that the calculation methods provide reliable results. Important quality assurance measures at the Federal Institute for Materials Research and Testing are given concerning the preparation, run and evaluation of a numerical analysis with reference to the appropriate guidelines. Hence, a successful application of the finite element (FE) method requires a suitable mesh. An analysis of the 1 m puncture drop test using successively refined FE meshes was performed to find an acceptable mesh size and to study the mesh convergence using explicit dynamic FE codes. The FE model of the cask structure and the puncture bar is described. At the beginning a coarse mesh was created. Then this mesh was refined in two steps. In each step the size of the elements was bisected. The deformation of the mesh and the stresses were evaluated dependent on the mesh size. Finally, the results were extrapolated to an infinite fine mesh or the continuous body, respectively. The uncertainty of the numerical solution due to the discretisation of the continuous problem is given. A safety factor is discussed to account for the uncertainty.
Transport of large nuclear power plant components - experiences in mechanical design assessment
(2010)
Mechanical design assessment approaches of actual spent fuel and HLW transport package designs
(2010)
Transport of large nuclear power plant components - experiences in mechanical design assessment
(2009)
Numerical simulation of 9 meter drop of a transport and storage cask with aluminium impact limiter
(2010)
For the purpose of numerical simulation of 9 meter drop of a transport and storage cask with aluminium impact limiter, an elastic-incremental plastic material model with strain rate hardening acc. to Cowper-Symonds is used for the development of isothermal as well as adiabatic stress-strain relations of aluminium from the compression test at constant ambient temperature. After that, two different simulation strategies are compared. At first, the drop test is calculated fully coupled, i.e. with isothermal stress-strain relations and possible heat generation in the material. Then the drop test is recalculated in a very simplified manner with adiabatic stress-strain relations from the compression test in an isothermal simulation. Both calculation strategies show similar results in the investigated load scenario.
BAM as competent authority regarding assessment of mechanical and thermal design, activity release and quality assurance aspects of SNF and HLW transport packages developed a thesaurus of experience feedback topics from recent approval procedures. The list is structured according to the European PDSR guide. It involves issues, which from BAM point of view, needed clarification during last package design assessment procedures. The list contains issues from operation (e.g. deactivation of handling lugs not intended for package handling), maintenance (e.g. leak tightness of impact limiting devices) to technical assessment (e.g. formation of hydrogen by radiolysis and its impact on pressure and ignitability, consideration of ageing mechanisms, thermo-mechanical assessment, impact of gaps between content and flask, spent fuel behaviour) as well as to general and specific safety analysis report requirements (e.g. report structure and required data).
Packages for the transport of radioactive material are often equipped with impact limiters consisting of wood, encapsulated by steel sheets. These impact limiters shall ensure the transport cask meets the mechanical and thermal IAEA regulatory test requirements. After damage caused by the mechanical tests the package has to withstand a severe fire scenario. According to the regulations during and following the thermal test, the specimen shall not be artificially cooled and any combustion of materials of the package shall be permitted to proceed naturally.
Corresponding to results of the French institute IRSN combustion and smouldering of wood inside the impact limiter occurred during and after the fire test. An additional energy supply from a pre-damaged impact limiter to the cask could be the consequence for the safety assessment of the containment.
BAM started a first test phase to examine the issue of combustion for such kind of package components. The goal was to understand the phenomena under the consideration of relevant regulatory boundary conditions. Several metal buckets were filled with wood and equipped with thermocouples. The test specimens have been prepared with different damage arrangements to take the influence of the mechanical tests into account. This paper shows the experimental setup and the conduction of the tests. The first test shows that pre-damaged metal encapsulations can lead to smouldering of the wood and with this to a supplement energy release after the end of the 30 minute fire. BAM is in the preparation process for a second test phase. A thermal test will take place with a wood filled test specimen weighing about 2Mg.
Storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage. Beside consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case. Not only the technical objects are subject of aging but also the safety demonstration basis is subject of “aging” due to possible changes of regulations, standards and scientific/technical knowledge. The basic document, defining the transport safety conditions, is the package design safety report (PDSR) for the transport version of the DPC. To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the PDSR, in connection with periodic renewals of package design approval certificates, have to be carried out. The main reviewing tool is a gap analysis. A gap analysis for a PDSR is the assessment of the state of technical knowledge, standards and regulations regarding safety functions of structures, systems and components.
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.