TY - CONF A1 - Weber, Mike A1 - Wille, Frank A1 - Ballheimer, Viktor A1 - Musolff, André T1 - From experiment to an appropriate finite element model-safety assessment for ductile cast iron casks demonstrated by means of IAEA puncture drop test N2 - 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. T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - IAEA KW - Puncture drop test KW - Numerical analysis KW - Verification PY - 2010 SP - 1-8 (Tuesday/T19/26-115) AN - OPUS4-23928 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kuschke, Christian A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Komann, Steffen T1 - Mechanical assessment criteria of spent fuel assemblies basket design N2 - 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. T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Transport conditions KW - Spent fuel assemblies KW - Basket design PY - 2010 SP - 1 EP - 7 (Thursday T36/92) AN - OPUS4-24345 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Weber, Mike A1 - Wille, Frank A1 - Ballheimer, Viktor A1 - Musolff, André T1 - From experiment to appropriate finite element model-safety assessment for ductile cast iron casks demonstrated by means of IAEA puncture drop test N2 - 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. KW - IAEA puncture drop test KW - Cylindrical cask KW - Explicit dynamics KW - Scaled model KW - Numerical analysis KW - Validation KW - Safety assessment PY - 2011 DO - https://doi.org/10.1179/1746510911Y.0000000010 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 22 IS - 3 SP - 148 EP - 153 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-25356 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Rolle, Annette A1 - Droste, Bernhard T1 - Mechanical behaviour of high burn-up SNF under normal and accident transport conditions - present approaches and perspectives N2 - 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. T2 - ASME 2012 Pressure vessels & piping division conference CY - Toronto, Ontario, Canada DA - 15.07.2012 KW - Fuel rods KW - High burn-up KW - Mechanical behaviour KW - Transport conditions PY - 2012 SP - PVP2012-78302, 1-7 PB - American Society of Mechanical Engineers CY - New York AN - OPUS4-26320 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Ballheimer, Viktor T1 - Mechanical Behaviour of High burn-Up SNF under Normal and Accident Transport conditions - Present Approaches an Perspectives - T2 - ASME Pressure Vessels & Piping Division Conference 2012 CY - Toronto, Canada DA - 2012-07-15 PY - 2012 AN - OPUS4-26326 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Apel, Andreas A1 - Ballheimer, Viktor A1 - Kuschke, Christian A1 - Schubert, Sven A1 - Wille, Frank T1 - Approach for the use of acceleration values for packages of radioactive material under routine conditions of transport N2 - 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. T2 - RAMTRANSPORT 2012 - 9th International Conference on the Radioactive Materials Transport and Storage CY - London, UK DA - 22.05.2012 KW - Radioactive materials KW - Safety assessment KW - IAEA regulations KW - Lost cases KW - Routine conditions of transport PY - 2012 SP - 1 EP - 8 AN - OPUS4-26747 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Quercetti, Thomas A1 - Ballheimer, Viktor A1 - Wieser, Günter T1 - Analytical, numerical and experimental investigations on the impact behaviour of packagings for the transport of radioactive material under slap down conditions T2 - PATRAM 2001, 13th International Symposium on Packaging and Transportation of Radioactive Materials CY - Chicago, IL, United States of America DA - 2001-09-03 PY - 2001 SN - 0893-6188 VL - 30 IS - 3 SP - 18 EP - 25 PB - Institute of Nuclear Materials Management CY - Northbrook, Ill. AN - OPUS4-1940 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zeisler, Peter A1 - Gogolin, Bernd A1 - Ballheimer, Viktor A1 - Pöppinghaus, J. T1 - Strain of Lid Bolts of Type B Packages During 9 m Drop Tests - Results of Real Tests and Problems of Analytical Predition T2 - PATRAM 2001 ; Annual International Symposium on Packaging and Transportation of Radioactive Materials CY - Chicago, IL, USA DA - 2001-09-03 PY - 2001 IS - CD-ROM PB - PATRAM CY - Chicago, Ill. AN - OPUS4-1883 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Neumann, Martin A1 - Wille, Frank A1 - Linnemann, Konrad A1 - Schubert, Sven A1 - Ballheimer, Viktor A1 - Sterthaus, Jens A1 - Droste, Bernhard T1 - Reflection on BAM mechanical design assessment of TN®24E spent fuel transport package N2 - 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. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 KW - Transport KW - RAM KW - Radioactive material cask KW - Nuclear fuel PY - 2013 SP - 1 EP - 12 PB - Omnipress AN - OPUS4-30270 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Ballheimer, Viktor A1 - Rolle, Annette A1 - Droste, Bernhard T1 - Mechanical behaviour of high burn-up SNF under normal and accident transport conditions - present approaches and perspectives - N2 - 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. T2 - PSAM 11 ESREL 2012 - 11th International probabilistic safety assessment and management conference & The annual european safety and reliability conference CY - Helsinki, Finland DA - 25.06.2012 KW - Spent fuel KW - High burn-up KW - Transport conditions KW - Mechanical analysis PY - 2012 SP - 1 EP - 9 AN - OPUS4-33850 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Komann, Steffen A1 - Neumann, Martin A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Weber, Mike A1 - Qiao, Linan A1 - Droste, Bernhard T1 - Mechanical assessment within type B packages approval: application of static and dynamic calculation approaches N2 - 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. KW - Package assessment KW - Package KW - Packaging KW - Structural analysis PY - 2011 DO - https://doi.org/10.1179/1746510911Y.0000000012 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 22 IS - 4 SP - 179 EP - 183 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-26448 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Qiao, Linan A1 - Neumann, Martin A1 - Komann, Steffen A1 - Ballheimer, Viktor T1 - Dynamic and quasi-static FE calculation of impact from 9 metre drop of spent fuel transport cask N2 - 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. KW - Drop test KW - Spent fuel transport cask KW - Finite element calculation KW - Dynamic simulation KW - Impact limiter KW - Wood PY - 2009 DO - https://doi.org/10.1179/174650909X12553336148026 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 20 IS - 4 SP - 174 EP - 178 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-20819 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Neumann, Martin A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Weber, Mike A1 - Qiao, Linan A1 - Droste, Bernhard T1 - Static and dynamic calculation approaches for mechanical design assessment of type B packages for radioactive material transport - 10193 T2 - WM2010 Conference CY - Phoenix, Arizona, USA DA - 2010-03-07 KW - Mechanische Bewertung KW - Radioaktives Material KW - Numerische Analyse KW - Typ B(U)-Verpackung PY - 2010 SP - 1 EP - 9 AN - OPUS4-21240 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Neumann, Martin A1 - Ballheimer, Viktor A1 - Wille, Frank T1 - Estimation of cask deceleration and impact limiter deformation under 9m drop test conditions using the calculation tool “ImpactCalc” T2 - PATRAM 2007, 15th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Miami, Florida, USA DA - 2007-10-21 KW - Impact limiter KW - Calculation KW - Maximum deceleration KW - Maximum deformation PY - 2007 SP - 1 EP - 8 PB - Institute of Nuclear Materials Management AN - OPUS4-17496 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Neumann, Martin A1 - Wille, Frank A1 - Ballheimer, Viktor T1 - Approximation of package deceleration and shock absorber deformation under 9 m drop test conditions using simplified numerical tool ImpactCalc KW - Impact limiter KW - Calculation KW - Maximum deceleration KW - Maximum deformation PY - 2008 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 19 IS - 1 SP - 35 EP - 40 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-17499 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Neumann, Martin A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Weber, Mike A1 - Qiao, Linan A1 - Droste, Bernhard T1 - Mechanical assessment within type B packages approval - the application of static and dynamic calculation approaches T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Mechanische Bewertung KW - Radioaktives Material KW - Numerische Analyse PY - 2010 SP - 1 EP - 8(?) AN - OPUS4-23384 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Neumann, Martin A1 - Günther, U. A1 - Ballheimer, Viktor A1 - Linnemann, Konrad A1 - Wille, Frank A1 - Droste, Bernhard T1 - Experiences by German authority with safety case approach for welding seams as part of SNF transport package containments N2 - 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. T2 - RAMTRANSPORT 2012 - 9th International Conference on the Radioactive Materials Transport and Storage CY - London, UK DA - 22.05.2012 PY - 2012 IS - Session 4.2 RAM 18 SP - 1 EP - 6 AN - OPUS4-26344 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Neumann, Martin A1 - Ballheimer, Viktor A1 - Sterthaus, Jens A1 - Schubert, Sven A1 - Eisenacher, Germar A1 - Wille, Frank T1 - Material characterization and modeling within safety analysis of packages for transport of radioactive material N2 - In mechanical design safety assessment ofpackages for the transport of radioactive material the International Atomic Energy Agency (IAEA) regulations [1,2] require the use of one of four methods of demonstration of compliance with regulations: tests with prototypes, reference to previous satisfactory demonstrations, tests with scaled models as well as calculation or reasoned argument. In all cases characterization of material behavior is necessary. The need for material characterization can be split into three groups: characterization for material modeling in the design and verification process, determination of assessment criteria and inspections during manufacturing. The paper deals primarily with material characterization for the material modeling in the design phase and authorities approval process. Main package components are discussed: cask body and lids, screws, trunnions, impact limiter and basket(structure holding the content in specified position). Depending on the material used for these components this paper Shows exemplary a possible set of material investigations in order to allow modeling in design and assessment during the approval process. T2 - ASME 2012 Pressure vessels & piping division conference CY - Toronto, Ontario, Canada DA - 15.07.2012 PY - 2012 SP - PVP2012-78297, 1-6 PB - American Society of Mechanical Engineers CY - New York AN - OPUS4-26271 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Zencker, Uwe T1 - Numerical approach to determine a package dependent bar length for the iaea pin drop test N2 - 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. T2 - PATRAM 2019 CY - New Orleans, LA, USA DA - 04.08.2019 KW - IAEA KW - 1-m-punch-bar-drop-test KW - Numerical approach KW - Bar length KW - Finite element analysis PY - 2019 UR - https://www.inmm.org/INMM-Resources/Proceedings-Presentations/PATRAM-Proceedings.aspx SP - 1 EP - 10 AN - OPUS4-49016 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Ballheimer, Viktor A1 - Neumeyer, Tino A1 - Wille, Frank T1 - Application of leakage rates measured on scaled cask or component models to the package containment safety assessment N2 - 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. T2 - 19th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM 2019 CY - New Orleans, LA, USA DA - 04.08.2019 KW - Leakage rate KW - Transport packages KW - Seals KW - Radioaktive PY - 2019 SP - Paper 1147, 1 AN - OPUS4-49053 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Ballheimer, Viktor A1 - Quercetti, Thomas A1 - Scheidemann, Robert A1 - Wille, Frank T1 - Design assessment by bam of a new package design for the transport of snf from a german research reactor N2 - 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. T2 - 19th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM 2019 CY - New Orleans, LA, USA DA - 04.08.2019 KW - Numerical modelling KW - Drop test KW - Assessment method KW - Ductile cast iron KW - Package design KW - Experimental testing PY - 2019 SP - Paper 1176, 1 AN - OPUS4-49054 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Ballheimer, Viktor A1 - Apel, Andreas A1 - Kuschke, Christian A1 - Moutarde, M. A1 - Desnoyers, B. A1 - Kalinina, E. A1 - Ammerman, D. T1 - ISO-Standard and IAEA guidance material for package load attachment Points - Current approaches and developments N2 - 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. T2 - 19th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM 2019 CY - New Orleans, LA, USA DA - 04.08.2019 KW - Load attachment KW - Stowage KW - Trunnion KW - Bolt design KW - Retention KW - Acceleration KW - Transport KW - Load cycles PY - 2019 SP - Paper 1130, 1 EP - 10 AN - OPUS4-49095 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Sterthaus, Jens A1 - Rolle, Annette A1 - Wille, Frank A1 - Vlassopoulos, Efstathios A1 - Papaioannou, Dimitrios T1 - Numerical Simulation of Spent Fuel Segments under Transport Loads N2 - 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. T2 - 17th International High-Level Radioactive Waste Management Conference (IHLRWM 2019) CY - Knoxville, Tennessee, USA DA - 14.04.2019 KW - Spent Nuclear Fuel KW - Finite Element Simulation KW - Transport packages PY - 2019 SN - 978-1-51088-669-8 SP - 1 EP - 7 AN - OPUS4-52046 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Quercetti, Thomas A1 - Scheidemann, Robert A1 - Komann, Steffen A1 - Ballheimer, Viktor A1 - Wille, Frank T1 - Package testing of a dual purpose cask for SNF from German research reactors N2 - 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. T2 - IHLRWM2019 CY - Knoxville, TN, USA DA - 14.04.2019 KW - Drop test KW - Package testing KW - Dual purpose cask PY - 2019 SN - 978-0-89448-761-3 VL - 2019 SP - paper 27283, 1 EP - 7 PB - ANS AN - OPUS4-50619 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -