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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. The cladding is the first barrier for the spent fuel pellets. An established method for characterising the cladding material is the ring compression test (RCT), in which a small, cylindrical sample of the cladding tube is subjected to a compressive load. Radial hydrides may precipitate in zirconium-based alloys (Zircaloy) under pre-storage drying and during slow cooling, which result in embrittlement of the cladding material and eventually a possible sudden failure of cladding integrity under additional mechanical loads. The focus of the presented research is on the development of appropriate nu-merical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. A modelling approach based on cohesive zones is ex-plained which is able to reproduce the propagation of cracks initiated at radial hydrides in the zir-conium 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 hy-drides 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. Based on the developed fracture mechanics approach, not only the deformation behaviour but also the fail-ure behaviour of irradiated as well as unirradiated Zircaloy claddings with radial hydrides under RCT loading conditions can be adequately described.
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. The cladding is the first barrier for the spent fuel pellets. An established method for characterising the cladding material is the ring compression test (RCT), in which a small, cylindrical sample of the cladding tube is subjected to a compressive load. Radial hydrides may precipitate in zirconium-based alloys (Zircaloy) under pre-storage drying and during slow cooling, which result in embrittlement of the cladding material and eventually a possible sudden failure of cladding integrity under additional mechanical loads. The focus of the presented research is on the development of appropriate nu-merical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. A modelling approach based on cohesive zones is ex-plained which is able to reproduce the propagation of cracks initiated at radial hydrides in the zir-conium 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 hy-drides 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. Based on the developed fracture mechanics approach, not only the deformation behaviour but also the fail-ure behaviour of irradiated as well as unirradiated Zircaloy claddings with radial hydrides under RCT loading conditions can be adequately described.
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 the Third Research Coordination Meeting on Spent Fuel Characterization, BAM as contributor to Task 3 reported on the activities and status of its research work.
Brittle Failure Limits of Spent Fuel Claddings Subjected to Long-Term Dry Interim Storage Conditions
(2024)
The mechanical properties of spent fuel claddings can be adversely affected under the conditions of long-term dry interim storage, so that the failure limits may be reached in case of mechanical loads during handling or transport after storage. Pre-storage drying and the early stage of interim storage can expose the cladding to higher temperatures and higher tensile hoop stresses than those associated with in-reactor operation and pool storage. During slow cooling of a cladding tube under internal pressure, radial hydrides may precipitate in zirconium-based cladding alloys. This can lead to embrittlement of the material and sudden failure of the cladding integrity under mechanical stress. In order to prevent brittle failure, numerical methods are being developed to predict the mechanical behaviour and identify limiting conditions. Experimental investigations, numerical analyses and evaluation methods are discussed.
An established experimental method for characterising cladding materials is the Ring Compression Test (RCT). Some test results on irradiated cladding tubes after operation in pressure water reactors are publicly available. However, it is helpful to carry out studies on unirradiated surrogate claddings with similar material properties to reduce the effort associated with irradiated samples in hot cells and to perform material tests with a wider range of parameters. On the basis of such experimental data, load-displacement curves have been numerically analysed for a selection of cladding materials.
Radial hydrides can cause a sample to break suddenly due to fracture even at low deformation. Noticeable load drops in the RCT are caused by 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 adjacent hydrides. The cohesive zone approach was used to simulate the failure process taking into account the radial hydride morphology.
The developed method can adequately describe both the deformation and failure behaviour of irradiated as well as unirradiated claddings of zirconium-based alloys with radial hydrides under RCT conditions. Limiting conditions can be expressed in terms of fracture energy and cohesive strength.
Brittle Failure Limits of Spent Fuel Claddings Subjected to Long-Term Dry Interim Storage Conditions
(2024)
The mechanical properties of spent fuel claddings can be adversely affected under the conditions of long-term dry interim storage, so that the failure limits may be reached in case of mechanical loads during handling or transport after storage. Pre-storage drying and the early stage of interim storage can expose the cladding to higher temperatures and higher tensile hoop stresses than those associated with in-reactor operation and pool storage. During slow cooling of a cladding tube under internal pressure, radial hydrides may precipitate in zirconium-based cladding alloys. This can lead to embrittlement of the material and sudden failure of the cladding integrity under mechanical stress. In order to prevent brittle failure, numerical methods are being developed to predict the mechanical behaviour and identify limiting conditions. Experimental investigations, numerical analyses and evaluation methods are discussed.
An established experimental method for characterising cladding materials is the Ring Compression Test (RCT). Some test results on irradiated cladding tubes after operation in pressure water reactors are publicly available. However, it is helpful to carry out studies on unirradiated surrogate claddings with similar material properties to reduce the effort associated with irradiated samples in hot cells and to perform material tests with a wider range of parameters. On the basis of such experimental data, load-displacement curves have been numerically analysed for a selection of cladding materials.
Radial hydrides can cause a sample to break suddenly due to fracture even at low deformation. Noticeable load drops in the RCT are caused by 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 adjacent hydrides. The cohesive zone approach was used to simulate the failure process taking into account the radial hydride morphology.
The developed method can adequately describe both the deformation and failure behaviour of irradiated as well as unirradiated claddings of zirconium-based alloys with radial hydrides under RCT conditions. Limiting conditions can be expressed in terms of fracture energy and cohesive strength.
A state-of-the-art (SOTA) review on characterisation of spent nuclear fuel (SNF) properties in terms of source term and inventory assessment (neutron, gamma-ray emission, decay heat, radionuclide inventory, elemental content) and in terms of out-of-core fuel performance (cladding performance and fuel integrity in view of the safety criteria for SNF interim storage, transport and canister packaging) using several numerical and experimental approaches and methodologies is presented. This SOTA report is a result of the spent fuel characterisation (SFC) work package as part of the European Joint Programme on Radioactive Waste Management (EURAD), which offers an overview of the status of knowledge in the field of SNF characterisation and assessment during the pre-disposal phase. The document aims to focus on the current safety-significant gaps and related challenges, providing a direct link to the goals of the mandated actors of EURAD. The report is expected to be used by all EURAD colleagues in their national programmes as a key resource for knowledge management programmes and to contribute to demonstrating and documenting the state-of-the-art.
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.
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.
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.