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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.
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.
In this work we present results for three typical sealing materials (EPDM, HNBR, FKM) after up to 100 days ageing at high temperature under hydrogen. To allow for a better evaluation of the occurring changes the results are compared with effects of ageing in air on the material properties. Despite the common assumption that ageing under hydrogen atmosphere is expected to be less severe for the material in com-parison to air ageing, this is not seen for every material.
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.
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.
Overview of ongoing research and future prospects on polyethylene neutron shielding materials at BAM
(2023)
The extension of the interim storage period of radioactive waste before disposal will cause additional challenges for the nuclear waste management in Germany, so that an extensive knowledge of the long-term performance of casks, including their components and inventories, will be required for future extended storage licenses.
Ultra-high and high molecular weight polyethylenes ((U)HMW-PE) are used for neutron shielding purposes in casks for storage and transport of spent fuel and high-level waste due to their extremely high hydrogen content. During their service life of several decades as cask components, the PE materials are exposed to neutron and gamma radiation from the radioactive inventory of the casks, mechanical assembling stresses and temperature. All these combined effects affect the material properties of such components which in turn may be crucial for some possible accident scenarios.
At the Bundesanstalt für Materialforschung und -prüfung (BAM), the effects of high temperature exposure in combination with subsequent or previous irradiation were investigated with a comprehensive aging program including thermal aging at 125 °C for different aging periods up to 5 years and irradiation with doses ranging from 50 to 600 kGy.
This contribution provides an overview of the ongoing research related to the structural changes of (U)HMW-PE induced by gamma irradiation and high temperature exposure and focuses on current research perspectives at BAM with regard to the prediction of the dynamic behavior of the material during extended interim storage in case of an accident scenario. First results of the coupled effect of temperature, radiation and mechanical loading will be presented. The effect of microstructural changes induced by gamma irradiation and high temperature on the mechanical behavior of (U)HMW-PE will be assessed.
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. Since the cladding is the first barrier for the spent fuel pellets, its integrity must be demonstrated until the end of interim storage and subsequent transportation. An established method for characterizing the cladding material is the ring compression test, in which a small, cylindrical sample of the cladding tube is subjected to a compressive load. This test is a laboratory representation of a load case where the fuel rod is crushed.
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 result in embrittlement of the cladding material and eventually a possible sudden failure of cladding integrity 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. Therefore, the study of hydride morphology plays an important role in describing the brittle failure behaviour of the claddings.
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. Typical hydride morphologies are shown. An iterative inverse analysis method is described 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 is able to 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.
This project as part of the European Joint Programme on Radioactive Waste Management has received funding from the European Union’s Horizon 2020 research and innovation programme under grant agreement no. 847593.
Elastomer seals are mounted as barrier seals in lid systems of containers designed for transport and disposal of negligible heat generating radioactive waste and as auxiliary seals in spent fuel storage and transportation casks (dual purpose casks (DPC)). When the behavior of mounted seals under normal and hypothetical accident conditions of disposal and transport is to be simulated, a comprehensive knowledge of their complex mechanical properties at every state of aging is necessary. In previous works, BAM’s efforts in experimental investigations on specimen artificially aged at different temperatures and times and the implementation of the found results in finite element material models were presented. Additionally, our approaches to reproduce the aging process itself and to extrapolate the results of artificially accelerated aging to longer times were presented. Numerical simulations have shown that the behavior of the seal during mounting and one-sided pressurizing and the resulting performance values such as leakage rate strongly depend on the coefficient of friction (COF) between flange and seal. The friction coefficient, in turn, depends on the aging state of the elastomer material as several publications suggest (see below).
Dynamic COF between an exemplary ethylene propylene diene rubber (EPDM) material and a stainless steel ball were determined by using a self-designed linear oscillation tribometer. Unaged and artificially aged EPDM specimen stored for 30 days and 100 days at a temperature of 150 °C were tested. A stainless steel ball (d=10 mm) is brought in contact with the specimen’s surface und loaded by normal forces of 2.5 N, 5 N, 10 N and 20 N. During a reciprocating movement of the EPDM sheet, the horizontal force/friction force is continuously measured, and the COF can be derived.
It is well known that friction is a complex phenomenon especially in soft materials. It cannot be excluded that the measured friction force is influenced by additional force components, resulting from the ball’s grooving through the elastomer’s surface. This force depends on the penetration depth of the ball and on the resistance of the elastomer in its different states of aging. The latter results from microstructural changes i.e., chain scission and additional crosslinking that occur during aging which in turn influence the softening or hardening of the material.
A finite element (FE) ABAQUS® model was developed to reproduce the measurement process. It should help to better understand the physical mechanisms and to quantify the percentage of measured forces resulting from real friction on the one hand and forces resulting from unintended side effects that could falsify the result on the other hand. The behavior of the elastomer in its different states of aging is reproduced by a FE material model already presented in previous works of BAM.
Hydrogenated nitrile butadiene rubber (HNBR) elastomer was thermo-oxidatively aged at different temperatures up to 150°C. Fourier transform infrared spectroscopy (FTIR), compression stress relaxation (CSR) and IRHD microhardness were used to characterise the chemo-mechanical changes of HNBR O-rings during thermo-oxidative ageing. FTIR shows the development of carbonyl, methyl and ester groups but the nitrile content was not affected by ageing. The effect of sample geometry during CSR was investigated. CSR data were converted through integrated kinetic laws. The conversion has proven its sensibility to detect heterogeneous ageing. This was confirmed by the IRHD measurements across the section of O-rings. Oxidation profiles were determined by applying the model of Wise et al. to the IRHD experimental results.
Suitable containers for receiving high-level radioactive waste are a key elements of every repository system concerning different host rocks like rock salt, claystone and crystalline rock as being considered in Germany. The disposal container design is decisive regarding boundary conditions for the transport, emplacement techniques, and the operational and long-term safety assessment. The systematic derivation of the requirements to be placed on disposal containers forms the basis for a targeted, comprehensible and transparent development of containers that meet the requirements. The presentation summarizes the outcomes of the research project KoBrA.
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 Second Research Coordination Meeting on Spent Fuel Characterization, BAM as leader of Task 3 reported on the status of the research work.
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 behaviour of spent nuclear fuel (SNF), irradiated cladding, fuel/cladding chemical interaction (FCCI) and ageing under conditions of extended interim storage, transportation and emplacement 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.
In order to demonstrate the safety of both spent fuel and the storage system, a good understanding of the processes that might cause deterioration is required. The International Atomic Energy Agency (IAEA) continued the coordinated research on Spent Fuel Performance Assessment and Research (SPAR) into a fourth phase starting in 2015 to evaluate fuel and materials performance under wet and dry storage and to assess the impact of interim storage on associated spent fuel management activities (such as handling and transport). BAM’s understanding and experiences in the field of long-term storage of spent fuel and especially spent fuel characterization are presented to assist the preparation of the final project report.
Embrittlement of spent fuel claddings during long-term dry interim storage - Current approach at BAM
(2019)
The integrity of fuel rods and cladding tubes must be demonstrated throughout the dry interim storage of spent nuclear fuel in Germany. For at least a relevant number of casks, an extension of interim storage between 40 and 100 years is foreseeable. The drying procedure for the fuel rods in conjunction with the thermomechanical storage conditions may change the microstructure of the cladding tubes and lead to their possible embrittlement. The failure probability might increase under mechanical load by handling procedures (transport after storage) on potentially embrittled cladding tubes of fuel rods especially at low temperatures after long-term storage. Results of ring compression tests on unirradiated pre-hydrided as well as irradiated samples with radial hydrides are investigated by finite element analyses to be able to describe the failure process and to identify failure criteria.
The characterization of spent fuel and radioactive waste is essential in the evaluation of the ability of these materials to meet acceptance criteria. Prior to determining a method for spent fuel characterization, it is necessary to identify the type of data required to support safety and other analyses for the storage, transportation and final disposal of spent fuel. BAM’s understanding and capabilities in the field of spent fuel characterization are presented for the scope of a coordinated research project on the topic.
A Memorandum of Understanding (MOU) between Argonne National Laboratory (Argonne) and the German Bundesanstalt für Materialforschung und -prüfung (BAM) was signed in October 2014. Its objectives are to promote cooperation among scientists and specialists at Argonne and BAM and establish a framework for collaboration in advanced technologies associated with the back end of the nuclear fuel cycle. Collaborative activities involving Argonne and BAM may be implemented through the promotion of joint research activities and scientific workshops and conferences; exchange of technical information; and visits by scientists, specialists, and graduate, postgraduate, and Ph.D. students. The MOU was renewed in January 2020 for another five years. Highlights of the progress and achievements are provided in identified topical areas for collaboration on ageing management guidance, storage, transportation and disposal R&D, joint conference activities, and conclude with the benefits of the MOU in fostering Argonne and BAM collaboration.
The general aim of this research is the development of methods for predicting mechanical behavior and identification of limiting conditions to prevent brittle failure of high-burnup (HBU) pressure water reactor (PWR) fuel cladding alloys. A finite element (FE) model of the ring compression test (RCT) was created to analyze the failure behavior of zirconium-based alloys with radial hydrides during the RCT. An elastic-plastic material model describes the zirconium alloy. The stress-strain curve needed for the elastic-plastic material model was derived by inverse finite element analyses. Cohesive zone modeling is used to reproduce sudden load drops during RCT loading. Based on the failure mechanism in non-irradiated ZIRLO® claddings, a micro-mechanical model was developed that distinguishes between brittle failure along hydrides and ductile failure of the zirconium matrix. Two different cohesive laws representing these types of failure are present in the same cohesive interface. The key differences between these constitutive laws are the cohesive strength, the stress at which damage initiates, and the cohesive energy, which is the damage energy dissipated by the cohesive zone. Statistically generated matrix-hydride distributions were mapped onto the cohesive elements and simulations with focus on the first load drop were performed. Computational results are in good agreement with the RCT results conducted on high-burnup M5® samples. It could be shown that crack initiation and propagation strongly depend on the specific configuration of hydrides and matrix material in the fracture area.