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Task 3 of Work package 8 (Spent Fuel Characterization) of the European Joint Programme on Radioactive Waste Management (EURAD) investigates the behaviour of nuclear fuel and cladding after discharge. The aim of the work is to understand and describe the be-haviour of spent nuclear fuel (SNF), irradiated cladding, fuel/cladding chemical interaction (FCCI) and ageing under conditions of extended interim storage, transportation and em-placement in a final disposal system. BAM contributes to the project as partner and leads Task 3. The presentation gives an overview of the project status, main achievements in experimental work and modelling studies, deviations from the plan, delays and challenges ahead.
The prevention of brittle fracture of spent fuel claddings during long-term dry interim storage is based on experimental investigations, numerical analyses and assessment methods for predicting the mechanical behavior and determining limiting conditions. The ring compression test (RCT) is an established experimental method for characterizing cladding material. Test results for various high-burnup pressure water reactor zirconium-based fuel cladding alloys (e.g., ZIRLO®, M5®) are publicly available. To reduce the effort associated with irradiated samples in hot cells, it is helpful to perform studies on unirradiated surrogate cladding material. Based on such experimental data, load-displacement curves were numerically analyzed for selected cladding materials. In the presence of radial hydrides, a sample may suddenly fail by fracture even at small deformations. Noticeable load drops in the RCT occur associated to unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting neighboring hydrides. The failure process was simulated by cohesive zones controlled by the fracture energy and the cohesive strength. A modeling approach is presented in which the radial hydride morphology is taken into account. Based on the developed fracture mechanics approach with cohesive zone modeling, not only the deformation behavior but also the failure behavior of irradiated as well as unirradiated claddings with radial hydrides under RCT loading conditions can be adequately described.
The presentation gives an overview about the current experiences of BAM regarding ageing management of transport packages in Germany.
The concept of the BAM-GGR 023 for ageing management of transport packagings will be described.
The fundamentals of BAM-GGR 023 will be shown, including the following key aspects:
-the organization of an Ageing Management System (AMS),
-the issue of an Ageing Management Plan (AMP),
-the ageing evaluation as a part of the AMP,
-the Ageing Surveillance Program (ASP) as a part of the AMP,
-the Ageing Management Documentation (AMD).
The storage of spent fuel and high level radioactive waste in Germany is performed in interim storage containers with double lid systems. The lids are bolted and equipped with metal seals (e.g. Helicoflex®) to ensure the safe enclosure of the inventory. The used metal seals have a layered structure consisting of three components as can be seen schematically in the cross-sectional view in Fig. 1.
In the center a helical spring is positioned that is surrounded by two C-shaped jackets and is mainly responsible for generation of the required restoring force. The inner jacket is made of stainless steel and homogenizes the restoring force of the helical spring. The outer jacket is made of silver or aluminium which both are soft metals in comparison to the contact partners (lid and container body). During bolting of the lid to the container body the seal is compressed. The generated restoring force of the helical spring causes a plastic deformation of the outer jacket and adapts to the surfaces of the lid and the container body. Hence, leakage paths are closed and the sealing function is generated.
Typical durations for existing interim storage licenses in Germany are 40 years, but it can be expected that they have to be extended to longer periods as a final repository will not be available before the end of the running licence periods. This extension of license periods requires a solid understanding of the long-term behaviour of the seals under storage conditions. To meet this challenge long-term investigations have been started at Bundesanstalt für Materialforschung und –prüfung (BAM) in 2009. These tests focus on seals assembled in test flanges which are stored at temperatures ranging from room temperature to 150 °C for accelerated ageing. The aged seals are tested repeatedly after certain ageing steps and the leakage rate as indicator for sealing performance, the remaining seal force, and the useable resilience upon decompression are determined.
In the poster an update on the performed investigations in respect to earlier publications (Grelle et al. 2019, Goral et al. 2023) will be given and the implications of the results for resilient long term safety will be presented. Additionally, a focus will be laid on the currently planned further investigations and the question “What is additionally needed for evaluation of an interim storage period extension in regard to the used metal seals?” will be addressed.
Beginning in 1976 the former iron ore mine Konrad was geologically investigated concerning its suitability as a location for a German final repository for low and intermediate level waste. After a comprehensive licensing procedure it was approved and is now being prepared and equipped for the planned start of storage in 2027. At the end of the 1980s, the requirements for waste containers for storage in this repository were defined for the first time on the basis of an incident analysis of the on-site handling procedures, beginning with the unloading of the packages after the arrival at the site up to the final positioning in the underground storage galleries. A 5-meter drop onto the rock ground of the mine was identified as the covering case for high mechanical requirements (ABK II container class). In contrast to the 9-meter drop according to the IAEA Regulations for the Safe Transport of Radioactive Material, the 5-meter drop is performed onto a hard but not essentially unyielding target. The container is typically not protected by an impact limiter. The requirements for a potential test facility are described in the regulations for the Konrad repository. The mechanical strength of the target is defined as a concrete strength equivalent to the identified properties of the rock ground of the mine.
Since 1991 BAM has consistently used precast reinforced concrete slabs as target for drop tests in the framework of licensing procedures as well as in research projects. While the original design fulfilled the requirement for the integrity of the concrete slab in most cases, it failed when drop tests onto an edge of a container were performed. A redesigned concrete slab developed in a research project and suggested as a reference target in 2009 has been successfully used in Konrad licensing procedures since then.
The paper gives a brief overview of the historical development and it describes BAM’s efforts and approaches to continuously guarantee a concrete slab of defined quality and to provide a test setup for valid drop tests from the applicants and the authorities view.
Beginning in 1976 the former iron ore mine Konrad was geologically investigated concerning its suitability as a location for a German final repository for low and intermediate level waste. After a comprehensive licensing procedure it was approved and is now being prepared and equipped for the planned start of storage in 2027. At the end of the 1980s, the requirements for waste containers for storage in this repository were defined for the first time on the basis of an incident analysis of the on-site handling procedures, beginning with the unloading of the packages after the arrival at the site up to the final positioning in the underground storage galleries. A 5-meter drop onto the rock ground of the mine was identified as the covering case for high mechanical requirements (ABK II container class). In contrast to the 9-meter drop according to the IAEA Regulations for the Safe Transport of Radioactive Material, the 5-meter drop is performed onto a hard but not essentially unyielding target. The container is typically not protected by an impact limiter. The requirements for a potential test facility are described in the regulations for the Konrad repository. The mechanical strength of the target is defined as a concrete strength equivalent to the identified properties of the rock ground of the mine.
Since 1991 BAM has consistently used precast reinforced concrete slabs as target for drop tests in the framework of licensing procedures as well as in research projects. While the original design fulfilled the requirement for the integrity of the concrete slab in most cases, it failed when drop tests onto an edge of a container were performed. A redesigned concrete slab developed in a research project and suggested as a reference target in 2009 has been successfully used in Konrad licensing procedures since then.
The paper gives a brief overview of the historical development and it describes BAM’s efforts and approaches to continuously guarantee a concrete slab of defined quality and to provide a test setup for valid drop tests from the applicants and the authorities view.
Beginning in 1976 the former iron ore mine Konrad was geologically investigated concerning its suitability as a location for a German final repository for low and intermediate level waste. After a comprehensive licensing procedure it was approved and is now being prepared and equipped for the planned start of storage in 2027. At the end of the 1980s, the requirements for waste containers for storage in this repository were defined for the first time on the basis of an incident analysis of the on-site handling procedures, beginning with the unloading of the packages after the arrival at the site up to the final positioning in the underground storage galleries. A 5-meter drop onto the rock ground of the mine was identified as the covering case for high mechanical requirements (ABK II container class). In contrast to the 9-meter drop according to the IAEA Regulations for the Safe Transport of Radioactive Material, the 5-meter drop is performed onto a hard but not essentially unyielding target. The container is typically not protected by an impact limiter. The requirements for a potential test facility are described in the regulations for the Konrad repository. The mechanical strength of the target is defined as a concrete strength equivalent to the identified properties of the rock ground of the mine.
Since 1991 BAM has consistently used precast reinforced concrete slabs as target for drop tests in the framework of licensing procedures as well as in research projects. While the original design fulfilled the requirement for the integrity of the concrete slab in most cases, it failed when drop tests onto an edge of a container were performed. A redesigned concrete slab developed in a research project and suggested as a reference target in 2009 has been successfully used in Konrad licensing procedures since then.
The paper gives a brief overview of the historical development and it describes BAM’s efforts and approaches to continuously guarantee a concrete slab of defined quality and to provide a test setup for valid drop tests from the applicants and the authorities view.
Cohesive Zone Modelling Approach on Irradiated Claddings Subjected to Long-Term Dry Interim Storage
(2023)
Long-term dry interim storage may adversely affect the mechanical properties of spent fuel rods, possibly resulting in a reduced resilience during handling or transport after storage. Pre-storage drying and the early stage of interim storage can subject the cladding to higher temperatures and higher pressure induced tensile hoop stresses than those associated with in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate in zirconium-based alloys (Zircaloy) during slow cooling, which may result in embrittlement of the cladding material and eventually a sudden failure of cladding under additional mechanical loads. Especially long, continuous radial hydride structures and low temperature can cause severe embrittlement of claddings and finally failure by fracture even at small deformations. The focus of the presented research is on the development of appropriate numerical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. An iterative inverse analysis method is used for deriving the elastic-plastic material properties in the hoop direction of a ring-shaped sample. A modelling approach based on cohesive zones is explained which can reproduce the propagation of cracks initiated at radial hydrides in the zirconium matrix. The developed methods are applied to defueled samples of cladding alloy ZIRLO®, which were subjected to a thermo-mechanical treatment to reorient existing circumferential hydrides to radial hydrides. A selected sample showing sudden load drops during a quasi-static ring compression test is analysed by means of fracture mechanics for illustrative purposes.
Acceptable limits for activity release from transport casks for high-level radioactive material specified in the IAEA regulations must be kept by the integrity of cask body and the cask sealing system. BAM as the German competent authority for mechanical, thermal and containment assessment of packages liable for approval verifies the activity release compliance with the regulatory limits. One of the fundamental aspects in assessment is the specification of conservative package design leakage rates.
To ensure the required package tightness for both, storage, and transport of the cask before and after storage usually metal seals of the Helicoflex® Type are used. Due to the long-term use the seal behavior is influenced by temperature and time. The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport specified in the regulations can have a significant effect on the leak tightness of the sealing system. Whereas the safety for application of new, non- aged Helicoflex® seals is verified sufficiently, there are still technical data gaps concerning the efficiency of aged Helicoflex® seals.
BAM performed experiments to learn more about the sealing efficiency of aged Helicoflex® seals with Aluminum and Silver outer jackets. The seals were compressed in test-flanges and for artificial ageing the complete flange systems were stored in an oven for several month at a high temperature. During the compression and decompression tests after the aging, load-deformation characteristics of the seals, and leakage rates were measured. With these tests a load situation was simulated, which can occur in the regulatory drop test of the cask: Under high impact loads the bolted lid can lift a little for a short moment, allowing a little movement of the seal, so that the contact area can change before compressing again.
The poster presentation will show details about test conditions and first results.
The European Joint Programme on Radioactive Waste Management (EURAD) is working on Spent Fuel Characterization (SFC) in its work package (WP) 8. Inspired by the EURAD activities, the International Atomic Energy Agency (IAEA) established an international Coordinated Research Project (CRP) on SFC. The EURAD WP SFC
participants are collaborating as a team on the IAEA CRP on SFC. The EURAD WP SFC project consists of four tasks. Task 3 investigates the behaviour of nuclear fuel and cladding after discharge. The aim of these activities is to understand and describe the evolution of the cladding-pellet system and its ageing under conditions of extended interim storage, transportation and emplacement in a final disposal system. At a Consultancy Meeting, BAM as contributor to Task 3 presented current results of the failure analysis of irradiated ZIRLO® claddings under conditions of the Ring Compression Test.
BAM acts as authority and for service in safety assessment of packages for transport and storage of radioactive materials. We offer extensive test capabilities and application of analytical methods for design verification and simulation for all types of packages for the transport and storage of radioactive materials according with the international IAEA Regulations for the safe transport and for national storage acceptance criteria. BAM operates several test facilities for drop and stacking testing, leak testing and thermal testing. The large drop test tower allows dropping full-scale specimens up to 200,000 kg in any drop orientation as requested. The comprehensive test facilities combined with long-term experience, newest equipment and measurement devices according to the latest state-of-the-art technology ensures realisation of complex test campaigns for package safety evaluation.
Beyond that, non-destructive and destructive material test devices and experts are available. Equipment and application of all kinds of typical measurement categories can be offered for testing campaigns.
In recent years we performed testing of full-scale type B package models with complex handling and preparation procedures. The results were contributed for different package design approval procedures. Type A packages mainly designed for medical related transport purposes, were continuously tested according to the transport regulations over recent years as well. Moreover, we work on research topics with relevance to package safety. The mechanical behaviour of lid closure systems under transport and storage conditions and the thermal behaviour of impact limiters were recently of special importance for the assessment competencies of BAM and were investigated under use of our test facilities.
The paper describes the test facilities and capabilities for package design safety evaluation at BAM and shows examples from our recent work.
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.
BAM acts as authority and for service in safety assessment of packages for transport and storage of radioactive materials. We offer extensive test capabilities and application of analytical methods for design verification and simulation for all types of packages for the transport and storage of radioactive materials according with the international IAEA Regulations for the safe transport and for national storage acceptance criteria. BAM operates several test facilities for drop and stacking testing, leak testing and thermal testing. The large drop test tower allows dropping full-scale specimens up to 200,000 kg in any drop orientation as requested. The comprehensive test facilities combined with long-term experience, newest equipment and measurement devices according to the latest state-of-the-art technology ensures realisation of complex test campaigns for package safety evaluation.
Beyond that, non-destructive and destructive material test devices and experts are available. Equipment and application of all kinds of typical measurement categories can be offered for testing campaigns.
In recent years we performed testing of full-scale type B package models with complex handling and preparation procedures. The results were contributed for different package design approval procedures. Type A packages mainly designed for medical related transport purposes, were continuously tested according to the transport regulations over recent years as well. Moreover, we work on research topics with relevance to package safety. The mechanical behaviour of lid closure systems under transport and storage conditions and the thermal behaviour of impact limiters were recently of special importance for the assessment competencies of BAM and were investigated under use of our test facilities.
The paper describes the test facilities and capabilities for package design safety evaluation at BAM and shows examples from our recent work.
BAM-GGR 023 was published in June 2022
BAM-GGR 023 gives guidance to applicants regarding ageing management for competent authority approved package designs
Ageing management with evaluation of ageing mechanisms according to para 613A of IAEA SSR-6 (Rev. 1) and their effects are part of the general management system.
Essential items are: AMS (systemic-related), AMP (package design-related) and AMD (package-related)
There are different requirements for organization and extent of ageing management depending on type of use
There are also different requirements for ageing evaluation depending on type of use for packaging/package, classification of components and their accessibility and replaceability
Wood is widely used in impact limiters of Transport casks for radioactive material. Encapsulated by an outer and inner Steel structure, spruce wood is offen applied in layers of altemating direction. The friction at the interfaces between these layers is of crucial importance for the impact and energy absorption e.g., at an accidental impact of a cask against a hard component.
In order to get detailed Information for corresponding numerical calculations, in this study the friction coefficient for the combinations wood-wood and wood-steel was measured in the temperature ränge between -40 °C and 90 °C according to the relevant stress conditions for such casks. Results show decreasing friction with increasing temperature, ranging from 0.43 at -40 °C to 0.22 for 90 °C for wood-steel combinations and from 0.3 at -40 °C to 0.24 at 90 °C to for a wood-wood combination.
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.
The consideration of ageing mechanisms is with integration of the new para 613A into IAEA SSR-6 (Rev. 1) now obligatory for the design of transport packages. In addition, para 809(f) requires 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. Para 503(e) requires that all packaging components and radioactive contents have been maintained during storage in a manner that all requirements specified in IAEA SSR-6 (Rev.1) and in the applicable certificates of approval have been fulfilled.
The evaluation of ageing mechanisms and their effects including monitoring are part of BAM’s authority assessment tasks related to the mechanical and thermal package design and quality assurance aspects. BAM has compiled a guideline for the implementation of ageing assessment and of the measures for ageing management of the approval procedure based on requirements of IAEA SSR-6 (Rev.1). The guideline is applicable only for packages requiring a competent authority approval.
The paper aims to describe the structure of the guideline and the general approach for ageing management requirements. The type and amount of measures for ageing management depend mainly on the use of the package and on the ageing effects for the component, which result from relevant ageing mechanisms during package operation time.
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). These measures shall ensure that the anticipated changes of the package design under consideration of ageing effects still complies with the design approval specification. Therefore, an ageing surveillance program (ASP) and, if necessary, a gap analysis program shall 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.
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.
Packages for the transport of radioactive materials shall fulfil the requirements of the IAEA regulations for the safe transport. The requirements define mechanical and thermal test conditions including criteria ensuring the package design’s ability to withstand severe accidents and provide a high level of technical safety. Different methods can be used for safety demonstration showing compliance with the regulations.
The central part of a safety demonstration which is presented in this paper was a comprehensive drop test program with a full-scale model of a transport package accompanied by pre- and post-test FE analyses. Using full-scale drop test models allow the benefit that similarity and scaling issues become a significant smaller issue, additional material investigations can be limited and analyses for transferring test results to the original package design are reduced. Additionally, experience for the future serial packaging manufacturing and handling procedures can be collected in a very early state of the design approval process. The pre-test finite element analyses derived and justified the drop test program consisting of several drop sequences with different drop orientations of the specimen. The performance and the results of the drop test sequences shows the manageability and the advantage e.g., in view of the direct availability of test results for the package licensing. On the other hand, the drop test performance shows the difficulties during handling and the need for additional equipment during preparation of the specimen.
The package presented was intended for the transport and storage of compacted radioactive waste from reprocessing of spent nuclear fuel assemblies - designed and applied for approval by the AGC consortium. The project ended in 2021. The package design was characterized by a cask body made of a forged thick stainless-steel shell, a bolted double lid system with metallic gaskets and wood filled shock absorbers at both ends. The total mass of the entire transport package including content was 120,000 kg, the total length was about 7000 mm and the diameter approximately
3000 mm, both measures include the shock absorbers.
The paper provides an insight into the performance of a full-scale drop testing campaign within the package safety evaluation and shows some selected test results.