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In practice usually thin-walled cylindrical silos are supported locally at the bottom rim of the shell. In this area a stress concentration may occur, which is responisble for buckling of the shell. In the currently applied codes, there is no information, how to analyze the load-bearing capacity of theselocally supported shells. In this paper, a new design formula is given for determining the load-bearing capacity.
In accordance with the IAEA transport regulations, the design of special form radioactive material (SFRM) shall resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval includes besides the program for physical tests (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance, and inspection.
SFRM source design plus management system shall ensure, that every specimen of the approved design is able to survive the severe mechanical and thermal tests at any time of its SFRM-working life.
Due to the long-term use of SFRM designs in most cases, the assessment of the source ageing behavior is an important aspect in the approval procedure. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. Besides of radioactive content, corrosion is a main factor for possible SFRM design degradation.
Although the IAEA Advisory Material SSG-26 already implies an indication of the need for considering ageing mechanisms, suitable amendments in the regulatory requirements of SSR-6 should be introduced to make the approval procedure more transparent and help to reduce rounds of questions by the authority. A supplementary requirement for considering of ageing mechanisms could be a helpful contribution to an international harmonization of the approval procedure.
This paper will describe major influencing factors to be considered to assess the ageing behavior of a SFRM design and will identify the need for a regulatory specification of a SFRM-working life as basis for the assessment of the SFRM design regarding time-dependent weakening. A proposal for an explicit requirement for consideration of ageing mechanisms in safety assessment of SFRM, which should be considered in the ongoing SSR-6 revision cycle, will be explained.
In accordance with the IAEA transport regulations, the design of special form radioactive material (SFRM) shall resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval includes besides the program for physical tests (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance, and inspection.
SFRM source design plus management system shall ensure, that every specimen of the approved design is able to survive the severe mechanical and thermal tests at any time of its SFRM-working life.
Due to the long-term use of SFRM designs in most cases, the assessment of the source ageing behavior is an important aspect in the approval procedure. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. Besides of radioactive content, corrosion is a main factor for possible SFRM design degradation.
Although the IAEA Advisory Material SSG-26 already implies an indication of the need for considering ageing mechanisms, suitable amendments in the regulatory requirements of SSR-6 should be introduced to make the approval procedure more transparent and help to reduce rounds of questions by the authority. A supplementary requirement for considering of ageing mechanisms could be a helpful contribution to an international harmonization of the approval procedure.
This paper will describe major influencing factors to be considered to assess the ageing behavior of a SFRM design and will identify the need for a regulatory specification of a SFRM-working life as basis for the assessment of the SFRM design regarding time-dependent weakening. A proposal for an explicit requirement for consideration of ageing mechanisms in safety assessment of SFRM, which should be considered in the ongoing SSR-6 revision cycle, will be explained.
In accordance with the IAEA transport regulations Special Form Radioactive Material (SFRM) is either an indispersible solid radioactive material or a sealed capsule containing radioactive material.
The design of special form radioactive material has to resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval shall include besides the required test program (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance and inspection. The specified quality assurance measures have to assure, that every specimen of the approved design is produced in the same verified quality and every specimen must be able to survive the severe mechanical and thermal tests at any time of its working life.
Due to the long-term use of SFRM the consideration of ageing is an important aspect in the approval procedure by BAM, the competent authority for SFRM approval in Germany. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. So, besides of radioactive content, corrosion is a main factor for possible design degradation.
This paper will describe major influencing factors to be taken into account to assess the ageing behavior of a SFRM design and will emphasize that there is a need for a regulatory specification of a SFRM-working life as basis for the aging evaluations.
The consideration of ageing mechanisms is with integration into the IAEA-requirements now obligatory for the design of transport packages. For packages intended to be used for shipment after storage the consideration of the effects of ageing mechanisms during storage in safety analyses and the implementation of corresponding instructions for operation and maintenance are required. Therefore, these packages have been maintained during storage in a manner that all requirements specified in the IAEA-requirements and in the applicable certificates of approval have been fulfilled.
BAM is the competent authority in Germany for evaluating the thermal and mechanical design of packages for the transport of radioactive materials. In this context, BAM also evaluates the aging mechanisms and aging effects occurring during the long-term storage period of such transport packages. BAM has developed a guideline for the implementation of ageing assessment and of the measures for ageing management into the approval procedure based on IAEA-requirements.
The goal of the presented paper is to introduce this guideline and the general approach for ageing management requirements and to describe the application to a specific package design. The type and amount of the ageing management not only depend on the use of the packaging, but also on the accessibility of the components that are considered in the ageing assessment. The different operating phases of a packaging are considered for the ageing assessment, such as loading, interim storage or transport.
The implementation of measures for ageing management is divided into three levels – systemic measures, package design related measures and documentation. The systemic measures are attributed to the general management system and define the whole activities for organization of ageing management like structure, responsibilities, documentation, reports and evaluation. The package design related measures are defined in an ageing management plan (AMP). Therefore, an ageing surveillance program (ASP) and, if necessary, a gap analysis program has to be developed. The ageing management documentation (AMD) ensures the continuous documentation of the compliance of a specific package to the approved package design, comprising mainly records resulting from operation and surveillance.
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.
Assessment experience on packages loaded with damaged spent nuclear fuel for transport after storage
(2018)
In 2017 the first German package approval certificate was issued for a dual purpose cask (DPC) design with encapsulated damaged spent nuclear fuel. At the Bundesanstalt für Materialforschung und -prüfung (BAM) a comprehensive assessment procedure was carried out with respect to the mechanical and thermal design, the containment design and quality assurance for manufacturing and operation. Main objective of this procedure was to verify the Package Design Safety Report (PDSR) fulfils the requirements according to the IAEA regulations SSR-6.
Until now only standard spent nuclear fuel assemblies were designated for interim storage and transports. Due to nuclear phase out in Germany all other kinds of SNF in particular damaged fuel has to be packed. Therefore specific requirements have to be considered in accordance with international experiences written in IAEA technical reports. In Germany damaged spent nuclear fuel (DSNF) needs a tight encapsulation with special encapsulations and clearly defined properties.
Due to the limited amount of DSNF these encapsulations are designed for storage and transport in existing packages. From the assessment experience it has been seen, corresponding PDSR need an extensively expansion to cover the design of these encapsulations and their influences on the package. Then such well-defined encapsulations can be handled like standard fuel assemblies. The main difference to standard package components is, encapsulations with permanent closure achieve their specified condition not after manufacturing but only during operation after loading and closing.
Thus specific handling instruction and test procedures are necessary especially for welding, where BAM is able to survey the quality of this first part of operation.
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.
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.
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.
Assessment of quality management for transport packages not requiring authority design approval
(2018)
The majority of transports of radioactive materials are carried out in packages which don’t need a package design approval of a competent authority. Low active radioactive materials are transported in such kind of packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well.
In Germany the decision to phase out nuclear energy leads to a strong demand for packages to transport low and middle active radioactive waste due to the dismantling and decommissioning of nuclear power plants. 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 of Type A.
For the packages of Type IP-2, IP-3 and Type A an assessment by the German competent authority is required for the quality management for the design, manufacture, testing, documentation, use, maintenance and inspection. In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure.
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.
The first German package design approval certificate for a dual purpose cask intended for loading with damaged spent nuclear fuel was issued recently. BAM as part of the competent authority system in Germany carried out a comprehensive assessment procedure with respect to the mechanical and thermal design, the release of radioactive material and the quality assurance aspects of manufacturing and operation. Packages for the transport and storage of radioactive material have been assessed by BAM for many years, thus the common assessment procedure is well-known and good practice. Up to now only SNF without defects or HLW with well-defined properties were designated for long-term Interim storage and transports afterwards. Due to Germany’s nuclear phase out all other kinds of spent nuclear fuel in particular damaged spent nuclear fuel shall be packed as well. Damaged spent nuclear fuel needs a tight closure with Special encapsulations and clearly defined properties in Germany. In addition, these encapsulations shall be long-term durable, because they are not accessible after loading in a packaging within periodical inspections. The main difference to Standard package components is that encapsulations with a permanent closure achieve their specified conditions not after manufacturing but only during operation, after loading and closing. To ensure compliance with the specific conditions, special measures for quality assurance are necessary during operation of each encapsulation, e.g. drying and sealing, which were assessed by BAM. The present paper gives an overview of the conducted assessment from BAM and point out the findings concerning to the special closure lid of the approved encapsulation, which is screwed and welded. A wide verification concept is necessary to show the specific tightness under transport conditions. Together with quality assurance measures during first operation steps these encapsulations with damaged spent nuclear fuel can be handled like standard fuel assemblies in approved package designs.
When storage of spent nuclear fuel or high level waste is carried out in dual purpose casks (DPC), the effects of aging on safety relevant DPC functions and properties have to be managed in a way that a safe transport after the storage period of several decades is capable and can be justified and certified permanently throughout that period. The effects of aging mechanisms (e.g. radiation, different corrosion mechanisms, stress relaxation, creep, structural changes and degradation) on the transport package design safety assessment features have to be evaluated. Consideration of these issues in the DPC transport safety case will be addressed. Special attention is given to all cask components that cannot be directly inspected or changed without opening the cask cavity, like the inner parts of the closure system and the cask internals, like baskets or spent fuel assemblies. The design criteria of that transport safety case have to consider the operational impacts during storage. Aging is not the subject of technical aspects only but also of intellectual aspects, like changing standards, scientific/technical knowledge development and personal as well as institutional alterations. Those aspects are to be considered in the management system of license holders and in appropriate design approval update processes. The paper addresses issues that are subject of an actual International Atomic Energy Agency TECDOC draft 'Preparation of a safety case for a dual purpose cask containing spent nuclear fuel'.
When storage of spent nuclear fuel (SNF) or high-level waste (HLW) is done in dual purpose casks (DPC), the effects of aging on safety relevant DPC functions and properties have to be managed in a way that a safe transport after the storage period of several decades is capable, and can be justified and certified permanently throughout that period. The effects of aging mechanisms (like e.g. radiation, different corrosion mechanisms, stress relaxation, creep, structural changes and degradation) on the transport package design safety assessment features have to be evaluated. The consideration of these issues in the DPC transport safety case will be addressed. Special attention is given to all cask components which cannot be directly inspected or changed without opening the cask cavity, what are the inner parts of the closure system and the cask internals, like baskets or spent fuel assemblies. The design criteria of that transport safety case have to consider the operational impacts during storage. Aging is not subject of technical aspects only, but also of 'intellectual' aspects, like changing standards, scientific/ technical knowledge development and personal as well as institutional alterations. Those aspects are to be considered in the management system of the license holders and in appropriate design approval update processes. The paper addresses issues which are subject of an actual IAEA TECDOC draft 'Preparation of a safety case for a dual purpose cask containing spent nuclear fuel'.
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.