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The majority of transports of radioactive materials are carried out in packages which don’t need a package design approval by a competent authority. Low-active radioactive materials are transported in such packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well. Decommissioning of NPP’s leads to a strong demand for packages to transport low and middle active radioactive waste. According to IAEA regulations the “non-competent authority approved package types” are the Excepted Packages and the Industrial Packages of Type IP-1, IP-2 and IP-3 and packages of Type A. For these types of packages an assessment by the competent authority is required for the quality assurance measures for the design, manufacture, testing, documentation, use, maintenance and inspection (IAEA SSR 6, § 306). In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure. Their regulatory level in the IAEA regulations is not comparable with the “regulatory density” for packages requiring competent authority package design approval. Practices in different countries lead to different approaches within the assessment of the quality assurance measures in the management system as well as in the quality assurance program of a special package design. To use the package or packaging in a safe manner and in compliance with the regulations a management system for each phase of the life of the package or packaging is necessary. The relevant IAEA-SSR6 § 801 requires documentary verification by the consignor concerning package compliance with the requirements.
Prüfung der qualitätssichernden Maßnahmen an Verpackungen püfpflichtiger Versandstücke durch die BAM
(2012)
Transport of Large Nucklear Power Plant Components - Experiences in Mechanical Desig Assessment -
(2012)
Ziel dieses Beitrages soll es sein, zum einen das Vorgehen bei der Erlangung einer Anerkennung zur qualitätsgesicherten Herstellung einer Verpackung für prüfpflichtige Versandstücke der zuständigen Behörde im Sinne des ADR Unterabschnittes 1.7.3 'Qualitätssicherung' darzustellen und zum anderen Probleme bei der Umsetzung der rechtsverbindlichen Vorschriften des Gefahrgutrechts aufzuzeigen.
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.
Federal Institute for Materials Research and Testing (BAM) is the competent authority for mechanical and thermal safety assessment of transport packages for spent fuel and high level waste in Germany. In context with package design approval of the new German high level waste cask CASTOR® HAW28M, BAM performed several drop tests with a half scale model of the CASTOR® HAW/TB2. The cask is manufactured by Gesellschaft für Nuklear Service mbH and was tested under accident transport conditions on the 200 tons BAM drop test facility at the BAM Test Site Technical Safety. For this comprehensive test program, the test specimen CASTOR® HAW/TB2 was instrumented at 21 measurement planes with altogether 23 piezo resistive accelerometers, five temperature sensors and 131 triaxial strain gauges in the container interior and exterior respectively. The strains of four representative lid bolts were recorded by four uniaxial strain gauges per each bolt. Helium leakage rate measurements were performed before and after each test in the above noted testing sequence. The paper presents some experimental results of the half scale CASTOR® HAW/TB2 prototype (14 500 kg) and measurement data logging. It illustrates the extensive instrumentation and analyses that are used by BAM for evaluating the cask performance to the mechanical tests required by regulations. Although some of the quantitative deceleration, velocity and strain values cannot be shown because of confidentially issues, they are provided qualitatively to illustrate the types of measurements and methodologies used at BAM.
Transport of Large Nuclear Power Plant Components - Experiences in Mechanical Design Assessment -
(2011)
BAM (Federal Institute for Material Research and Testing) is the competent authority for mechanical and thermal safety assessment of transport packages for spent fuel and high level waste (HLW) in Germany. In context with package design approval of the new German HLW cask CASTOR HAW28M, BAM performed several drop tests with a half-scale model of the CASTOR HAW/TB2. The test model was manufactured by GNS (Gesellschaft fur Nuklear Service mbH) and tested under accident transport conditions on the 200 tons BAM drop test facility at the BAM Test Site Technical Safety. For this comprehensive test program the test specimen CASTOR HAW/TB2 was instrumented at 21 measurement planes with altogether 23 piezo-resistive accelerometers, five temperature sensors and 131 tri-axial strain gauges in the container interior and exterior, respectively. The strains of four representative lid bolts were recorded by four uniaxial strain gauges per each bolt. Helium leakage rate measurements were performed before and after each test in the above noted testing sequence. The paper presents some experimental results of the half-scale CASTOR HAW/TB2 prototype (14,500 kg) and measurement data logging.
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.