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Eingeladener Vortrag
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eavy-weight packages for the safe transport of radioactive material are equippedwith impact limiters often built ofwood-filled steel sheet structures to fulfil the requirements of the International Atomic Energy Agency (IAEA) regulations. The requirements definemechanical tests followed by a thermal test, including criteria ensuring the package design’s ability to withstand severe accidents and provide a high level of technical safety. Impact limiters are a package component mainly designed for the packages to withstand severe mechanical accident scenarios. In drop tests the impact limiters absorb the kinetic energy during impact of the package. The package must then - with its pre-damaged impact limiters - endure a thermal test defined precisely in the IAEA regulations as a 30-min fully engulfing 800 °C-fire. After the fire, a wood-filled impact limiter may continue to release thermal energy from an ongoing combustion process, defining relevant package temperatures. The energy flow from a possible burning impact limiter to the package is important for the safety evaluation of transport packages. To investigate the combustion behaviour of densely packed layers of spruce wood, encapsulated in pre-damaged cylindrical metal enclosures, a test set-up has been realised. The set-up consists of a combustion chamber to perform these tests under defined boundary conditions. The temperature development of the test specimens will be observed fromoutside with a thermographic imager, with HD-Cameras, and the mass loss will be measured during the entire test. Airflow conditions in the combustion chamber are analysed using Computational Fluid Dynamics (CFD) calculations in OpenFOAM. The planned combustion test setup is described.
Packages for the transport of radioactive material are designed to en-dure severe accidents. Packages for the transportation of radioactive material must demonstrate that the package can withstand certain prescribed tests from the IAEA Regulations [1]. In addition to mechanical tests, a thermal test in form of a fire test must be carried out. As packages to be tested at BAM are signifi-cantly larger than previous package designs, BAM has expanded an existing fire test stand. A modular concept is chosen, which means that the arrangement of the burner nozzles can be adapted to the test specimen. The dimensions of the burner rings, the type, the orientation and the number of burner nozzles can be varied depending on the test specimen. In addition, various pumps can be used to set the corresponding mass flow. With the help of a calorimeter test, the fire test stand can be qualified for a specific size of packages regarding the boundary conditions of the IAEA Regulations [1]. Due to the typically wood filled impact limiters in German package designs, a fire test is necessary, as experiments have shown that possible openings that occurred during a mechanical test contributed to the igni-tion of the wood filled impact limiters within the prescribed 30 minutes of the IAEA Regulations [1]. From a series of experiments, two experiments are pre-sented to show the possibilities to obtain different temperatures and temperature rates in the test specimen. In addition to sensor data, the heat flux into the package is calculated to verify that the IAEA boundary conditions are satisfied. In addition to the temperature data, other data such as wind speed and wind direction are also recorded to explain subsequent effects in the measurement data in a comprehen-sible manner.
Germany had been transporting spent fuel to the reprocessing plant in Sellafield and La Hague for decades until around 2005. Resulting from the obligation to take back the vitrified high-level waste from reprocessing six CASTOR® HAW28M filled with 168 canisters with radioactive waste in vitrified form were transported from Sellafield to the interim storage facility Biblis in 2020. Rail wagons were used for the transport to the port in Barrow-in-Furness, where they were loaded into a dedicated seagoing vessel, certified as INF Class 3 according to the INF Code. This was the first time that vitrified high level waste with considerable heat load was transported under a German design approval certificate.
BAM was involved in the authority assessment of the conditions for the sea transport. For the first transport BAM required among others, assessment of temperature distribution during transport, logging of temperatures of cargo bays and graphical imaging of temperatures of the bay with the cask in order to ensure compliance with temperature specifications, e.g. maximal neutron absorber and gasket temperatures. Special interest was taken in the identification of possible events exceeding the specified temperatures considering the different philosophies of IMDG code and its supplement INF code regarding temperature control of hatches. Results show compliance with assumed conditions.
In einem laufenden Kooperationsprojekt zwischen der BAM Berlin und der MPA Stuttgart wird das probabilistische Master Curve (MC) Konzept nach ASTM E1921 hinsichtlich seiner Übertragbarkeit auf ferritisches Gusseisen mit Kugelgraphit (DCI) untersucht. Zielsetzung ist die Bereitstellung einer Vorgehensweise zur Bestim-mung und Bewertung der dynamischen Bruchzähigkeit von DCI im Übergangsbe-reich. In diesem Beitrag wird der Versuchswerkstoff der Festigkeitsklasse GJS-400 näher vorgestellt. Ferner werden erste Ergebnisse von zwei bruchmechanischen Versuchsserien an SE(B)140-Großproben bei Belastungsraten von ca. 6∙104 MPa√ms-1 bei -40 °C und 60 °C präsentiert.
Within fracture mechanics safety assessment for steels in nuclear technology, the probabilistic master curve (MC) concept according to ASTM E1921 is currently used for quasi-static loading conditions as a supplement to the established deterministic ASME reference curve concept. However, for ferritic ductile cast iron (DCI), a systematic review of potential modifications to the assumptions and the procedure according to ASTM E1921 and an associated validation are still lacking. For this reason, the application of the fracture mechanics MC concept to ferritic ductile cast iron is being investigated in a joint research project between MPA Stuttgart and BAM Berlin. According to IAEA SSG-26, the safety assessment of DCI containers for transport as well as storage of radioactive materials is based on the fracture mechanics criterion of general crack initiation exclusion. Within this context, the focus of the research project is on brittle fracture. The goal is the establishment of a methodology to determine and assess dynamic fracture toughness values of DCI in the ductile-to-brittle transition regime, using samples extracted from a component.
The experimental program uses a DCI material of grade GJS-400, which is generally used for transport and storage containers of radioactive materials. Firstly, a basic mechanical-technological material characterization will be performed consisting of tensile, Charpy and Pellini tests. Following the material characterization, the loading rate relevant for brittle fracture will be determined using instrumented C(T)25-fracture tests at a temperature of -40°C. To be able to transfer the relevant loading rate to other specimen geometries or sizes, the time-dependent course of the Weibull stress is assessed via numerical analyses. With this information an extensive experimental program consisting of DC(T)9, C(T)25, C(T)50, SE(B)10, SE(B)25 and SE(B)140 specimens is performed. These experimental results will be further supported by numerical and fractographic analyses. The empirical experimental database developed for DCI materials will be used to assess the applicability of the MC concept for DCI materials under dynamic loading.
At the current time manufacturing and precracking of test specimens is ongoing. The authors provide first results concerning the mechanical-technological material characterization, the relevant loading rate for brittle fracture determined via C(T)25 tests, and some early results of other specimen types. These experimental results are also supported by preliminary numerical and statistical analyses.
Ultrasonic echo testing is a well-established non-destructive testing technique used to investigate the interior of concrete structures in civil engineering. For the reconstruction of internal features, methods similar to seismic imaging are applied. To improve the ultrasonic imaging capabilities, we investigate the potential of a newly available 3D matrix mode measurement device in combination with advanced imaging techniques.
Commonly used ultrasonic array measurement devices rely on several ultrasonic transducers coupled to act as one transducer that transmits or receives transversal ultrasound waves. Although this approach is well suited to detect elongated structures such as pipes and rebars, it has limitations if 3D embedded objects are present. Hence, the ultrasonic device MiraA1040Pro used here is designed to measure the ultrasonic wavefield on a 4 by 16 matrix of individual transducers. For testing, we used laboratory specimens with linear and spheric embedded structures first. Both linear and matrix mode measurements were conducted and analysed using the Synthetic Aperture Focusing Technique
(SAFT) which is similar to Kirchhoff migration. Second, we used a data set from Teutschenthal mine at a shotcrete specimen constructed to simulate engineered barriers tailored for nuclear waste repositories. Ultrasonic measurements were investigated as a tool for quality assurance of these structures. The specimen contains both artificial 3D artifacts as well as naturally occurring defects such as a delamination. To make use of the advanced data acquisition, focusing 3D imaging techniques are applied to further improve the imaging quality.
Generally, the reconstructed images from the 3D matrix mode data have a lower level of Signal-to-Noise-Ratio than the 2D linear mode because the source signal is weaker compared to the linear mode. Moreover, complex wave conversions of the SH-transversal wave occur that are reduced in the 2D linear scenario. However, clearer images can be achieved at non-linear features, particularly with focusing imaging methods. Thus, combining 3D data acquisition techniques with advanced imaging methods improves the success in imaging complex concrete structures. This is of particular interest for thick concrete structures in nuclear barrier systems or foundations. Additionally, we demonstrate opportunities to use well-constrained test laboratory scenarios from non-destructive testing as a practical test case for geophysical methods.
Evaluation of Heat Fluxes in Fire Reference Test Conducted in BAM Propane Gas Fire Test Facility
(2024)
Packages for the transport of intermediate- and high-level radioactive waste are designed to withstand severe accidents. The International Atomic Energy Agency (IAEA) has established specific mechanical and thermal tests. Packages for the transport of radioactive material must withstand these tests to comply with the Regulations for the Safe Transport of Radioactive Materials IAEA [IAEA (2018)]. A fire reference package was developed with the primary objective to demonstrate that the fire meets the regulatory requirements. Another aim is to characterise the boundary conditions of the actual fire as input parameters for thermo-mechanical simulations. A simple method to characterise the boundary conditions of a real steady state fire with a fire reference package is presented. The thermal test mainly consists of a 30 minute fully engulfing 800°C pool fire or an equally severe fire, such as a propane gas fire. The fire reference tests are performed prior to the actual fire test with the real package. The heat fluxes into the package depend substantially on the fire characteristics and the surface temperature of the package. To investigate local and overall heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for repeated use. The fire reference package presented in this paper represents the outer geometry of a small transport container for radioactive material and is used as a device in civil engineering. It is designed as a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 182 mm and a diameter of 102 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material. Open-air fire tests were performed in a BAM propane gas fire test facility with the fire reference package. The measured temperatures are used to determine the changes of heat fluxes into the fire reference package in relation to the package surface temperature. The calculated heat fluxes allow its fitting to express the thermal exposure as simple mathematical boundary condition. Therefore, in a first approach, fire properties such as adiabatic surface temperature (AST) as proposed by Wickström et al. (2007), convection coefficient and emissivity are determined mathematically fitting the heat flux development presented in this paper. The evaluated results provide an initial picture of local fire characteristics of the conducted propane gas fire and are a further development of previous works from Feldkamp et al. (2020). The results can be used in thermal and thermo-mechanical models to simulate the load on the real transport package in fire. The test shows that the examined propane gas fire covers the IAEA-fire over a wide range of surface temperatures with the chosen test stand parameters.
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.
Within fracture mechanics safety assessment for steels in nuclear technology, the probabilistic master curve (MC) concept according to ASTM E1921 is currently used for quasi-static loading conditions as a supplement to the established deterministic ASME reference curve concept. However, for ductile cast iron (DCI), a systematic review of potential modifications to the assumptions and the procedure according to ASTM E1921 and an associated validation are still lacking. For this reason, the application of the fracture mechanics MC concept to ferritic ductile cast iron is being investigated in a recently started joint research project between MPA Stuttgart and BAM Berlin. According to the IAEA guidelines SSG 26, the safety assessment of DCI containers for transport as well as storage of radioactive materials is based on the fracture mechanics criterion of general crack initiation exclusion. Within this context, the focus of the current research project is on the further development of brittle fracture safety assessment methods. Therefore, a methodology shall be established to determine and assess dynamic fracture toughness values of DCI in the ductile-to-brittle transition regime, using samples extracted from a component. For the experimental program, a DCI material will be produced as it is generally used for transport and storage containers. A basic mechanical-technological material characterization consisting of tensile, Charpy and Pellini tests is planned and a comprehensive fracture mechanics test program was compiled covering dynamic testing of DC(T)9, C(T)25, C(T)50, SE(B)10, SE(B)25, and SE(B)140 specimens. Numerical analyses of different specimen geometries and test setups are planned to investigate the influences of stress triaxiality and loading rate. Throughout the experimental program, the microstructure-property relationship will be investigated based on quantitative metallographic and fractographic analyses. Finally, different MC assessment strategies will be considered and modifications will be made if necessary.
In dem Vortrag werden der Forschungsansatz, die Ziele, das Untersuchungsprogramm und die erwarteten Ergebnisse des BMUVForschungsvorhabens MCGUSS "Untersuchung des Master Curve-Konzeptes für ferritisches Gusseisen mit Kugelgraphit" vorgestellt. Mit Blick auf bruchmechanische Sicherheitsbewertungen in der Kerntechnik werden Stähle momentan mit dem deterministischen ASME Referenzkurvenkonzept, ergänzt durch das probabilistische Master Curve-Konzept nach ASTM E1921, bewertet. Das Master Curve-Konzept wird bereits für die Bewertung von ferritischen Schmiedestahlbehältern verwendet.
Bewertungsmethoden für ferritisches Gusseisen mit Kugelgraphit sind im Moment unvollständig. Normen wie die nationale BDG-Richtlinie P300 oder der internationale ASME-Code, Sect. III, Div. 3, WC-2330 decken Sprödbruch nicht ausreichend ab. Die Verwendung des Master Curve-Konzepts für ferritisches Gusseisen mit Kugelgraphit würde eine bruchmechanische Bewertung via einer KJc,d Verteilung erlauben und gleichzeitig Größeneffekte und einen Temperaturbereich von 100 K berücksichtigen.
The safe storage of radioactive material in nuclear waste repositories is a major task of our societies. The sealing function of such repositories is depending on the ability of the applied construction materials to form and maintain a geotechnical barrier with the surrounding rock. Corrosion of the construction material can lead to leakage of radioactive compounds and must therefore be avoided or minimized. Since concretes based on conventional Portland cement have not been found suitable in this context, alternative binders must be evaluated.
One such class of binders are hybrid cements, which are blends of low fractions of Portland cement or Portland clinker and high fractions of supplementary cementitious materials, such as blast furnace slag and fly ash, activated by an alkali salt. Besides a low heat of hydration and a sufficient early compressive strength [1], chemical similarities with ancient Roman concrete suggest an excellent durability in saline environments [2], which makes these cements potentially suitable for applications in nuclear waste repositories in evaporite rock, such as the Morsleben repository in Germany.
In the present study, two previously designed and characterized [3] hybrid cements, an alkali-activated slag/fly ash blend, and an OPC paste were studied regarding their resistance against corrosion in an aggressive saline solution. The saline solution was designed by the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) [Association for Facility and Reactor Safety] as a reference representing a solution forming as the result of contact of surface water with evaporite rock. The cement pastes were exposed to the saline solution up to 70 days and characterized by X-ray diffraction, thermogravimetric analysis and spatially resolved X-ray fluorescence spectroscopy. In addition, thermodynamic modelling was performed to simulate the alterations of the phase assemblage with increasing exposure to the saline solution and, thus, provide indications on the long-term durability of the cement pastes.
The experimental results revealed a correlation between the Portland clinker content of the cements and the resistance of the cement pastes against attack by the saline solution. This outcome was related to the formation of portlandite when sufficient clinker was available, which maintained the pore solution pH at ~12.5, and thus prevented the dissolution of cementitious phases. Once portlandite was consumed, C-N-A-S-H and ettringite dissolved and released calcium, aluminium, and hydroxide ions in the solution, maintaining the pH at ~10. In this pH range, the formation Cl-AFm phases was observed. At lower pH values, i.e., extended exposure durations, gypsum was the major corrosion product. The pH-dependent dissolution and formation of phases lead to pronounced zonation in the exposed cement pastes. Thermodynamic modelling indicated that the dissolved silicon from C-N-A-S-H reacts with magnesium ions in the saline solution to form M-S-H, and that long-term exposure eventually leads to a material rich in amorphous silica and brucite.
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).
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.
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.
Im Rahmen des Seminars "Angewandte Forschungsthemen (SAFT)" am Institut für Geowissenschaften der Christian-Albrechts-Universität zu Kiel werden verschiedene geophysikalische Methoden vorgestellt, die bei der zerstörungsfreien Prüfung im Bauwesen zum Einsatz kommen. Neben einem Überblick der Arbeiten an der BAM im Allgemeinen und dem Fachbereich im Besonderen geht es vor allem um den Einsatz von Ultraschallmethoden. Drei wesentliche Schwerpunkte werden dabei beleuchtet:
1) (Geophysikalische) Antworten auf Herausforderungen bei Ultraschall-Anwendungen
2) Material research and multi-sensory monitoring for concrete sealing structures in rock salt unterground repositories (Projekt SealWasteSafe)
und 3) Untersuchung von Spritzbeton-Probekörpern unter Tage mit Ultraschall zur Qualitätssicherung von Verschlussbauwerken.
Insgesamt werden diverse Möglichkeiten für die Geophysik für Forschung und Studierende im Bereich der zerstörungsfreien Prüfung im Bauwesen aufgezeigt.
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
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 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.
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.
Most transports of radioactive materials are carried out with packages not requiring competent authority approval of design. These encompass – in accordance with the IAEA SSR-6 [1] regulations – packages of the classification excepted, Industrial packages Type 1, 2 and 3 and Type A packages. Currently an upsurge in number and variation of these package designs can be seen in Germany, resulting from the phase out of nuclear energy in Germany as well as e. g. increased use of radioactive material for medical purposes.
A design assessment regarding the package safety is required in the international IAEA SSR-6 [1] regulations. BAM operates facilities for the performance of all regulatory tests required such as drop towers for a wide range of package masses and dimensions, fire test, leak tightness measurements and pressure test facilities. Experiences with several package types are shown.
Additionally, IAEA SSR-6 [1] requires the establishment of a management system for design, manufacture, maintenance, and repair of the packaging as well as for the preparation, consigning, loading, carriage, unloading and receipt of the package. Relevant for Germany, BAM has published guidance material on the process of management system acceptance in the technical guide BAM-GGR 016 [2]. The requirements encompass quality management plans for the manufacturing of packages including independent manufacturing surveillance and specific instructions for operation, maintenance, and repair of packagings. Examples for management system specifics and requirements are given.
Since the latest edition of the IAEA SSR-6 [1] regulations an ageing evaluation including systematic ageing management measures are required for all kind of package types. BAM is going to update the guidance material BAM-GGR 016 [2] to support the stakeholders with relevant information to fulfil the ageing aspect for packages not requiring competent authority approval. The paper explains how the ageing aspect may be included in the safety evaluation process and the management system measures and will give an outlook for the future guidance material.
The safety demonstration of heavy weight type B transport packages used for storing spent nuclear fuel (SNF) or radioactive waste is ensured by a combination of physical testing and numerical calculations. While experiments are performed in accordance with the IAEA regulations for selected drop scenarios, Finite-Element-Method (FEM) simulations are used to predict the most damaging case and to investigate additional drop positions.
BAM as competent authority in Germany has performed different investigations of a welding seam for a typical large transport package made of A508 forged steel, where the bottom plate is welded to the cylindrical shell. The package has a mass of approx. 120 t. Results of physical drop tests with a full-scale model and accompanying preliminary FEM simulations are presented to determine the decisive stresses in the welding seam. A drop test only represents one set of a package and test parameters. A further parameter analysis is considered to account for allowable variations of packaging properties (e.g. resulting from the manufacturing process) and, based on IAEA requirements, the temperature dependence of the material behaviour. The results of the stress analyses from the drop test and the simulation form the basis and provide the input parameters for a fracture mechanics analysis. In addition to the IAEA specifications, further standards are taken into account for an in-depth investigation, see R6 [1], BS 7910 [2] and API 579-1/ASME FFS1 [3].
All the above-mentioned standards require a manufacturer-specific defect analysis with respect to size and position. Both result from the welding process and the following heat treatment regime. The maximum defect sizes are ensured with non-destructive test methods (such as ultrasonic or particle methods) as integral part of the manufacturing process of the welding seam. Another important parameter in the welding process is the residual stress (secondary stress). The combination of the primary and secondary stress determines the total stress in the welding seam. The most damaging case of the welding seam is determined and evaluated with help of the above-mentioned standards and taking into account the IAEA requirements with respect to defect sizes, material properties, primary and residual stress, yield strength etc.
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.
Within the project SealWasteSafe, we advance construction materials and non-destructive monitoring concepts of sealing structures applied for underground disposal of nuclear waste. As these engineered barriers have high demands regarding structural integrity, an innovative alkali-activated material (AAM) that is highly suitable for the application in salt as a host rock is improved and tested on two laboratory scales. This AAM has a low heat evolution due to the reaction mechanism in comparison to common salt concretes based on Portland cement or magnesium oxychloride binders. Hence, crack formation due to thermally induced stress during the hardening process is reduced.
After successful laboratory tests with small specimens (height ~5 cm), comparably manufactured large cubic (edge length 70 cm) and cylindrical specimens (height 120 cm, diameter 40 cm) are equipped with sensing technologies to demonstrate the sensors´ technical capabilities. A comprehensive multi-sensory monitoring scheme is developed and investigated to characterize and compare the different material behaviour during the setting and hardening process of two materials: (1) the newly developed AAM-based mortars with salt aggregate, and (2) a blended Portland cement-based salt concrete as reference. The analysed parameters include temperature and humidity of the material, acoustic emissions, and strain variations recorded by fiber optic cables. Passive sensor systems based on radiofrequency identification technology (RFID) embedded in the concrete provide an interface for the wireless readout of various sensors. In parallel to the embedded RFID sensors, conventional cabled systems to read out the temperature and humidity measurements are installed for comparison. Additionally, a detailed inspection of the two large cubic specimens after a monitoring period of more than six months has been undertaken. Active thermography and ultrasonic echo measurements are used to reveal potentially occurring inner cracks from the surface. To verify the non-invasive results, a core sample (diameter 2 cm) was extracted from each of the investigated cubic specimens and analysed in detail with X-ray computed tomography.
Furthermore, ultrasonic methods are used for quality assurance to detect obstacles, cracks, and delamination at in-situ scale sealing structures. Experimental layout and applied imaging techniques are optimised to enhance the image quality for measurements from the front side of the engineered barrier. To characterize the inside of the test sealing structure and to improve the detection of potentially existing cracks, an ultrasonic borehole probe using the phased array technique is developed. First analyses at a half-spherical specimen coincide with modelling results and prove the reliability of the directional response caused by the phased array technique of the newly constructed ultrasonic borehole probe. Overall, the project SealWasteSafe helps to characterize construction materials and improves multi-sensory monitoring concepts and ultrasonic equipment for the sake of quality assurance. Particularly for salt as a host rock, this will help to design safe sealing structures for nuclear waste disposal.
A small cylindrical fire reference package was designed. A fire test setup was created using the fire reference package in accordance with IAEA standards. Four propane gas fire tests were conducted to determine heat fluxes into the fire reference package. The initial test setup was modified to create the final design of the fire test facility. This allowed for the heat flux to be adjusted to meet the regulatory IAEA fire qualification criteria. Furthermore, a numerical model of the fire reference package was created using boundary conditions derived from the experimental data. The simulation results demonstrated good agreement with the experimental data and provided additional insights.
Für die sichere Endlagerung nuklearer Abfälle sind im Salzgestein Verschlussbauwerke aus Salzbeton als geotechnischen Barrieren vorgesehen. Um die Integrität der Bauwerke im Rahmen einer Qualitätssicherung nachzuweisen, eignet sich eine zerstörungsfreie Prüfung (ZfP) mit Ultraschall-Verfahren. Potenziell auftretende Risse oder Ablösungen sowie Einbauteile können detektiert werden. Um die Möglichkeiten und Grenzen der Ultraschall-Echo-Messungen sowie der Auswertemethoden zu untersuchen, wurden in einem aus Spritzbeton hergestellten Testbauwerk in der Grube Teutschenthal (bei Halle, Deutschland) verschiedene Störkörper während der Betonage eingebettet. Die Störkörper (Breite 3 cm bis ~8 cm) liegen in Tiefen von etwa 30 cm bis 80 cm und repräsentieren Hohlräume, Risse oder Inhomogenitäten des Betons. Von der Oberfläche führten wir Ultraschall-Echo-Messungen auf einer Fläche von etwa 100x150 cm² in einem 3D Modus durch. Die Auswertung erfolgte zunächst mithilfe der in der ZfP etablierten Synthetic-Aperture-Focusing Technique (SAFT), die der Kirchhoff Migration ähnelt. Insgesamt wird eine Vielzahl von Reflektoren im Bauwerk abgebildet, die jedoch nur in wenigen Fällen den eingebauten Störkörpern zugeordnet werden konnten, so dass auch die Grenzen des Verfahrens bei dieser Bauweise deutlich werden. Allerdings konnte eine ungeplante großflächige (>30 cm) Ablösung nachgewiesen und durch anschließende Bohrungen verifiziert werden. Um das strukturelle Abbild des Bauwerksinneren trotz des stark streuenden Spritzbetons zu verbessern, wenden wir fokussierende Migrationsverfahren an. Erste Ergebnisse zeigen, dass eine klarere Abbildung einzelner Reflektoren erreicht werden kann. Dies erleichtert die Detektion einzelner Objekte. Das hier präsentierte Experiment mit bekannten Reflektoren unter realen Bedingungen mit Spritzbeton unter Tage bietet hervorragende Möglichkeiten zur Überprüfung von methodischen Weiterentwicklungen im Bereich der Ultraschallanalyse. Durch die Verbesserung der Auswertemethodik und der darauf basierenden ultraschallbasierten Qualitätssicherungskonzepte werden insgesamt sicherere Verschlussbauwerke für nukleare Endlager ermöglicht.
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.
Multi-sensory monitoring and ultrasound for quality assurance at underground sealing structures
(2023)
Within the safety concepts of underground disposal sites of nuclear waste, engineered barriers play an important role. As these sealing structures have high demands concerning integrity, we aim at advancing the available construction materials, monitoring and inspection techniques within the project SealWasteSafe.
After successful laboratory analysis on the cm-scale, cylindrical and cuboid specimens on the m-scale are produced from classical salt concrete and a specifically developed alkali-activated material. A comprehensive multi-sensory monitoring scheme is applied to compare the setting process of both materials and to demonstrate the sensors’ resistance to highly alkaline environments. Besides temperature and humidity of the materials, strain variations using fibre optic sensors and acoustic emissions are recorded over periods of at least 28 days, partly for more than 8 months. Passive sensor systems based on radiofrequency identification technology (RFID) embedded in the concrete allow for wireless access to temperature and humidity measurements and are compared to conventional cabled systems. After hardening and removal of the outer casings of the specimens, further non-destructive evaluations using ultrasonic echo and thermographic measurements are conducted. Preliminary results clearly highlight differences between the tested materials, particularly showing lower acoustic emission activity for the newly developed alkali-activated material potentially indicating less phase changes or cracks.
Complementary, ultrasonic methods are improved to be used for quality assurance to detect obstacles, potential cracks and delamination at in-situ sealing structure scale. A unique large aperture ultrasonic system (LAUS) with depth penetration as large as 9 m has already successfully been applied at the test site in Morsleben, Germany, of the federal company for radioactive waste disposal (BGE). Modelling studies help to further optimize the measurement layout. Advanced imaging techniques applied to the modelled and measured data will further improve the obtained images of internal structures. Additionally, an ultrasonic borehole probe is developed and constructed using phased arrays to further enhance the detection of potential cracks. Modelling and preliminary results from laboratory specimens prove the feasibility and potential of the directional response even in heterogeneous material such as concrete. Final investigations under in-situ conditions at the test site of the sealing structure are planned.
Overall, the project SealWasteSafe improves the construction material, multi-sensory monitoring, and ultrasound for quality assurance to allow for the development of safe nuclear sealing structures. Although the techniques are tailored for sealing structures within salt as a host rock, they are transferrable to a wider field of applications and alternative disposal conditions.
During the webinar "Dismantling and waste management: a playground for innovation" the innovative aspects of the PREDIS work package "innovations in cemented waste handling and pre-disposal storage" are presented. The webinar is jointly organised by Women in Nuclear Decommissioning (WINDI) and Women in Nuclear Innovation (WINI). The shown innovations include quality assurance (mainly NDE) and monitoring technologies with associated sensor systems, digital twin technology, and appropriate data handling including a decision framework. An overview of the on-going research in the field of pre-disposal waste management in the project Predis (WP7) is summarised.
The structural integrity of sealing structures in underground repositories for radioactive waste is a major concern regarding safe disposal. Non-destructive testing plays a very important role in the safety of this sealing structures. Ultrasonic pulse echo technology has been established in the construction industry as a non-destructive testing method for locating potential defects and material characterization of concrete structures. A borehole probe is an ultrasonic pulse-echo device used to investigate the internal structures of salt concrete. A prototype borehole probe is successfully demonstrated with the existing research borehole in the test site using a commercially available flaw detector. A novel phased array borehole probe that allows the beam to be focused based on the delay law was constructed and tested. The directivity measurements on the semi-cylindrical concrete sample agree well with the simulation results for the different focusing angles. In combination with geophysical reconstruction methods, the optimized radiation pattern of the probe is expected to improve the signal quality and thus increase the reliability of the imaging results. In addition, a unique large aperture ultrasound system (LAUS) is expected to provide enhanced images of possible cracks and embedded objects in in-situ structures. LAUS will be optimized to reliably detect objects at depths greater than 9 m and, at the same time, provide good resolution. Modeling studies show the potential to increase the distance between the transducer arrays further, and the same configuration has been successfully tested on a thick concrete wall. Overall, the paper shows that quality assurance using ultrasonic testing is crucial for constructing safe sealing structures to dispose of radioactive or toxic waste.
The friction at the interfaces between pieces of wood and steel is of crucial importance for the impact and energy absorption in impact limiters of transport casks for radioactive material. Here, the friction coefficient for the combinations wood-wood and wood-steel was measured in the temperature range between -40°C and 90°C. 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 for a wood-wood combination.
This presentation summarizes the opbjectives and the progress of the EURATOM project PREDIS (pre-disposal management of radioactive waste) and its workpackage 7 in particular. Focus is on the development of wireless sensors for monitoring radioactive waste packages, measuring radiation, temperature, pressure and humidity.
Für die Beförderung von umschlossenen radioaktiven Stoffen (Strahler) können Transporterleichterungen gelten, wenn das Strahlerdesign nachweislich unfallsicher ausgelegt ist und eine Zulassung als radioaktiver Stoff in besonderer Form vorliegt. Die in Deutschland zuständige Behörde für die Prüfung und Zulassung radioaktiver Stoffe in besonderer Form ist die Bundesanstalt für Materialforschung und –prüfung (BAM).
Eine oft langfristige Nutzung der Strahler kann eine alterungsbedingte Schwächung der Umschließung des radioaktiven Stoffes zur Folge haben. Jeder Strahler einer zugelassenen Bauart muss jedoch im Beförderungsfall zu jeder Zeit, auch nach längerer Nutzung, den vorgeschriebenen Prüfanforderungen genügen. Die Alterungsbewertung auf der Grundlage einer vom Antragsteller zu spezifizierenden Nutzungsdauer des radioaktiven Stoffes in besonderer Form ist seit langer Zeit Bestandteil des Zulassungsverfahrens in Deutschland. Ein von Deutschland eingebrachter Vorschlag im Rahmen der anstehenden Revision der IAEA Empfehlungen für die sichere Beförderung von radioaktiven Stoffen (SSR-6) soll die Lesbarkeit des Regelwerkes bezüglich der Alterungsbewertung verbessern, Klarheit über die damit verbundenen Anforderungen erzeugen und damit zu einer Harmonisierung der Zulassungsverfahren auf internationaler Ebene beitragen. In diesem Beitrag sollen wichtige Einflussfaktoren auf das Alterungsverhalten von radioaktiven Stoffen in besonderer Form aufgezeigt und die Notwendigkeit für die Spezifikation einer Nutzungsdauer als Grundlage für die Alterungsbewertung begründet werden. Der deutsche Vorschlag für die Revision des IAEA Regelwerks zum sicheren Transport radioaktiver Stoffe (SSR-6) wird vorgestellt und erläutert.
With our experiments we want to simulate a load situation which can result from a cask vertical drop test. The lid can lift a little for a very short moment and that can perhaps result in a little movement of the seal, a little rotation or a little lateral movement, so that the contact section can change. The aim of the project is to learn more about the sealing efficiency after the seal is compressed again. What we know from former component tests with non-aged seals is, that the leakage rate measured after one or more decompression /compression cycles with seal movement becomes significant higher. And this result is considered for specification of design leakage rates for release calculation in safety cases.
In the recent years we all have learned more about the significant influence of temperature and time on seal behavior and so the plan was to do similar experiments with aged seals, just to get an idea about seal behavior and achievable leakage rates.
What we did in detail is:
We compressed Helicoflex-seals of both design types, with Aluminum and Silver outer jackets, in testflanges and aged them in an oven at a temperature of 125°C for 3 months to produce seal properties comparable with properties after several years use at more typical operating temperatures between 90 and 100°C.
After this ageing procedure we opened the flanges completely, moved the seal a little to vary the contact area and compressed the flanges again. What we measured during all compression and decompression cycles was the load, the deformation and -as long as possible- the leakage rates.
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.
Untersuchung einer Schweißnaht mit verschiedenen Codes. Vergleiche von experimentellen und numerischen Ergebnissen
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.
Underground sealing structures are necessary to seal radioactive or toxic waste in underground repositories. Particularly developed MgO concrete is used in the mine in Teutschenthal and applied at a large-scale test with the aim to provide concrete with low permeabilities. The test structures (up to 10 m long) were created with the help of a shotcrete procedure. Besides destructive tests, non-destructive ultrasonic measurements are used for quality assurance to monitor potential anomalies such as cracks, concrete inhomogeneities, and delamination. We show results of ultrasonic measurements with a multi-transducer system used at the front site of the concrete structure. Images are obtained by synthetic aperture focusing techniques. Concreting sections are not systematically imaged so that a successful concreting is assumed as also indicated by observed low permeabilities. Several reflectors up to a depth of 1.2 m are identified and interpreted as potential damages in the concrete. Phase analysis of prominent reflectors reveal a negative impedance contrast thus indicating air filled voids, cracks or delamination. Boreholes through some of the identified reflectors are used to verify the results from ultrasonic measurements. Additionally, an experiment with incorporated defects is undertaken to analyse the reliability of the applied methods. Overall, the conducted tests show the great potential of ultrasonic measurements to detect critical anomalies. Despite challenges at small-scale structures (cm-order), large-scale anomalies can be identified. Implementing ultrasonic monitoring during and after the construction of concrete sealing structures is recommended as a tool for quality assurance.
Falltürme sind Bauwerke, die als Versuchsanlagen sehr speziellen, impulsartigen Belastungen ausgesetzt sind. Am Fallturm auf dem Testgelände Technische Sicherheit (TTS) der BAM wurde bei Routineinspektionen eine Abnahme der strukturellen Integrität in Form von Vorspannungsverlusten in den Bolzen der Stahlverbindungen festgestellt. Um ein Verständnis für die zugrundeliegenden Trag- und Schädigungsmechanismen zu erlangen, wurde ein umfassendes Structural Health Monitoring (SHM) System geplant und am Bauwerk installiert unter Nutzung von digitalen Bauwerksmodellen. Für die Auslegung des Überwachungssystems, insbesondere aber zur Unterstützung der Untersuchung des Schädigungsprozesses, wurden Finite-Elemente-Modelle erstellt. Um sicherzustellen, dass die FE-Modelle das reale Tragwerksverhalten mit ausreichender Genauigkeit abbilden, mussten sie jedoch in Bezug auf die gemessenen Antworten des Tragwerks kalibriert werden. Der vorliegende Beitrag beschreibt experimentelle und numerische Untersuchungen zur Identifizierung des strukturellen Systems des Stahlrohrgitterturms in Vorbereitung einer Überwachungskampagne. Die Auswertung von gemessenen Schwingungen unter ambienter Anregung ermöglichte die Identifizierung der Eigenfrequenzen mehrerer globaler Schwingungsmoden des Fallturms. Zur Modellvalidierung wurde zunächst eine Sensitivitätsanalyse durchgeführt, um die Parameter mit dem größten Einfluss zu ermitteln. Anschließend wurde ein evolutionärer Algorithmus (EA) zur Optimierung nach dem Prinzip der Minimierung der Differenzen zwischen gemessenen und simulierten charakteristischen Antworten eingesetzt. Das aktualisierte Modell wurde schließlich an der dynamischen Reaktion der Turmstruktur infolge einer realen Falltest-induzierten Stoßbelastung validiert. Die Ergebnisse zeigten eine gute Übereinstimmung zwischen numerischen und experimentellen Ergebnissen.
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.
Experimental and numerical investigation of prestressed bolt connections under lateral displacements
(2022)
The containment system of transport packages for spent nuclear fuel and high-level waste usually includes bolted lids with metal gaskets. The packages are assessed to specific transport conditions which are specified in the IAEA safety standards SSR-6 (IAEA 2018). These transport conditions, especially the so-called accident conditions of transport, imply high dynamic loading on the lids and the bolt connections of the package. The response of the lid systems on the mechanical accident conditions is generally investigated by experimental drop tests or numerically, e.g., by finite element analyses. The interpretation of the drop test results for the verification of the numerical models is often not obvious due to the complex superposition of different effects in the real tests. BAM has started a research project to get a better understanding about the behavior of prestressed bolt connections under loadings typical for these drop tests. In this context an experimental test set-up was developed to investigate the response of a single bolt connection under a prescribed lateral displacement of clamped parts. The bolt is instrumented by strain gauges to get the pretensional, the torsional and the bending stress in the bolt shank. Furthermore, the lateral movement and the tilt of the bolt head is measured during the test. A finite element model of the test set-up has been created in Abaqus FEA (Simulia 2021). The very detailed instrumentation of the test set-up shall give the opportunity to investigate and validate the numerical model. The aim of this paper is to give an overview about the proposed research project and to present first results.
Neue SCO-III Regularien für den Transport grosser Gegenstände als oberflächenkontaminierte Objekte
(2022)
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.
The latest 2018 edition of the International Atomic Energy Agency (IAEA) Regulations for the Safe Transport of Radioactive Material, No. SSR-6 [1], incorporates regulations for the shipment of large objects as a new category of surface contaminated object, SCO-III, based on the IAEA “performance package” concept. This paper provides background and practical guidance on these regulations. Additionally, the experiences of BAM with the appoval of two steam converters of the NPP Lingen are presented as the first approval process for SCO-III objects in Germany.
The primary additions to SSR-6 include SCO-III classification and requirements, approval and administrative requirements for the new classification, and the addition of SCO-III to the proper shipping name for UN 2913. Advisory material drafted for the new requirements will be included in the next revision of SSG-26, Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, expected to be published soon. Note that at this time the proposed provisions for large objects do not include components such as reactor vessels, due to the more limited experience and greater radioactivity levels. The SCO-III concept lays the groundwork and may be extended to cover other large objects that are classified as low specific activity (LSA) material in the future.
Transport packages shall satisfy various safety criteria regarding mechanical, thermal and radiation phenomena. Typical requirements focusing mechanical aspects are usually drop tests in accordance with IAEA regulations [1]. The drop tests are usually carried out experimentally and, as an additional measure, finite element analyses (FEA) are performed. A specific part of the investigations presented is the evaluation of the welding seam connecting cask shell and cask bottom. Experimental results and FEA are presented and compared. The evaluation of the welding seam performed includes a variety of aspects. In addition to the experimental and analytical stresses determined, different standards are used to investigate a possible crack initiation. Several destructive and non-destructive tests are performed for quality assurance in the manufacturing process as well as for different input parameters. The necessary monitoring and non-destructive measurement methods to define the boundary conditions of the standards are introduced. Taking into account all required parameters, the welding seam is examined and evaluated using the failure assessment diagrams (FAD) of the respective standards. It can be shown under the given boundary conditions that considering the experimental data, the welding seam is in the context of crack initiation below the enveloping curve in the acceptable region. More critical drop tests to be conducted are proposed and need to be investigated in future work.
The project "SealWasteSafe" of the Bundesanstalt für Materialforschung und -prüfung (Berlin) deals with sealing structures applied for underground disposal of nuclear waste from two perspectives: material improvement for application in sealing constructions and feasibility study regarding multi-sensor approaches to ensure quality assurance and long-term monitoring.
One specimen of 150 l made of alkali-activated material, which was found innovative and suitable for sealing constructions based on preliminary laboratory studies, and another one made of conventional salt concrete, are manufactured with an integrated multi-sensory setup for quality assurance and long-term-monitoring. The specimens were left in their cast form and additionally thermally insulated to simulate the situation in the repository. The multi-sensory concept comprises RFID technology embedded in the specimens suppling material temperature and moisture measurements, integrated distributed fibre optic sensing allowing strain measurement and acoustic emission testing for monitoring crack formation. Overall, the suitability and the functionality of the sensors embedded into and attached to strongly alkaline (pH > 13 for the AAM) and salt corrosive (NaCl) environment was proven for the first 672 h.
First RFID measurement succeeded after 626 h for the alkali-activated material and after 192 h for the conventional salt concrete. Strain measurement based on fibre optic sensing turned out the alkali-activated material (> 1 mm/m) undergoing approximately twice the compression strain as the salt concrete (< 0.5 mm/m). In contrast, the acoustic emission first and single hits representing crack formation in numbers, was found for alkali-activated material half of that detected at the salt concrete.
Neue Ultraschall-Messsysteme (LAUS, Bohrlochsonde) haben ihre Funktionstüchtigkeit unter Beweis gestellt. Ergebnisse aus den Messungen konnten verifiziert werden. Standard-Mess- und –Auswerteverfahren müssen für Spritzbeton weiterentwickelt und angepasst werden. Mit sämtlichen Messmethoden konnten großflächige Risse detektiert werden. Messungen mit der Bohrlochsonde und LAUS sind zur Qualitätssicherung geeignet.
Since cladding acts as the first barrier for nuclear fuel, it is key to understand the degrading mechanisms leading to cladding embrittlement. Zirconium alloy based cladding samples subjected to simulated pre-storage drying conditions have shown high susceptibility to brittle failure under the occurrence of hoop stress and low temperature as expected after long-term interim storage. The stress state associated with a compressive pinch-load scenario is reproduced by the Ring Compression Test (RCT), which results in sudden failure at relatively small mechanical loads with high sensitivity to the presence of radial hydrides. As part of the BAM research project BRUZL (Fracture mechanics analysis of spent fuel claddings during long-term dry interim storage), static RCTs were conducted on non-irradiated, pre-hydrided ZIRLO® ring-shaped samples that were subjected to a radial hydride treatment. All samples failed in macroscopically brittle fashion with sudden load drops and severe crack propagation through almost the entire wall thickness. In a post-RCT examination of the fracture surfaces using scanning electron microscopy, radial hydrides could be identified as the weakest link of the structure. However, the microscopic failure mechanism is much more complex due to cracks initiating at different axial positions in depth direction of the ring, overlapping cracks, and arresting cracks. A considerable portion of the fracture surfaces appeared to have undergone plastic deformation of the zirconium matrix resulting in void growth and coalescence associated with ductile failure. The geometry of the ductile patches was measured, and the ductile surface fraction was determined. Based on these observations, a statistical model was developed that can reproduce the composition of the fracture surface as a one-dimensional projection. A micromechanically motivated cohesive zone model (CZM) was implemented in existing RCT finite element analysis (FEA) models with different constitutive laws for brittle hydrides and ductile matrix respectively. In this work, FEA results are presented for non-irradiated ZIRLO® and high-burnup M5® claddings. The impact of varying hydride-matrix combinations is discussed.
Acknowledgement:
The project was funded by the former Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561.
Es werden Ergebnisse von 1H-NMR-Relaxometrie-Messungen an Salzbeton und einem alkali-aktivierten Material, die im Rahmen des Themenfeldprojektes Seal Waste Safe durchgeführt werden, vorgestellt. Dabei wird die zeitliche Entwicklung von ausgewählten Merkmalen der ermittelten T2-Relaxationszeitenverteilungen während der Hydratation betrachtet. Als Referenz zur Identifikation von charakteristischen Hydratationsereignissen dienen der Vicat-Nadeleindringversuch und Wärmeflusskalorimetrie.
Nach einer kurzen Einführung in die Schallemissionsprüfung und -Analyse wird ein Versuchsstand vorgestellt, der im Rahmen des Themenfeldprojektes Seal Waste Safe installiert wurde. Schwerpunkt des Vortrages sind die Schallemissionsmessungen an einem 150 l Demonstrator aus Salzbeton und einem weiteren aus alkali-aktiviertem Material. Neben der konventionellen Schallemissionsanaylse mit Aktivitäts- und Intensitätsparametern der Zeitsignale, werden auch frequenzbasierte Parameter zur Analyse andiskutiert.
Investigation of the internal impact during a 9 m drop test of an accident-safe waste package
(2021)
The safety assessment of packages for the transport of radioactive material follows the IAEA regulations and guidance. The specified regulatory tests cover severe accidents and demonstrate the package containment system integrity. Special attention must be drawn to the behaviour of the content which could move inside the package due to unpreventable gaps caused by the loading procedure and the structure of the content. A possible internal impact of the content which occurs during the drop tests onto the lid system is investigated. The IAEA regulations SSR-6 and the Guidance SSG-26, revised recently, consider input from Germany and France related to the significance of internal gaps.
In the context of a waste package design assessment, a model was equipped with a representative content to conduct a drop test with an internal impact. The weight and kinetic impact of this content covered all possible real contents. The objective of the test was to maximize the load onto the lid system and to prove the mechanical integrity by complying with the required leak tightness. The test was conducted conservatively at a package temperature lower than -40 °C at the BAM Test Site Technical Safety.
This paper gives an overview of efforts to address internal gaps and their consequences, and the BAM efforts with the implementation of this topic into IAEA regulations and guidance material. The paper then focuses on the conduction of a drop test and investigation of internal component impact.
A novel ultrasonic borehole probe is developed for the quality assurance of sealing structures in radioactive waste repositories using existing research boreholes. The aim is to analyse the sealing structures made of salt concrete concerning potential cracks, delamination, and embedded objects. A first prototype of the probe uses 12 individual dry point contact (DPC) shear horizontal wave transducers separated by equidistant transmitter/receiver arrays, each consisting of six individual transducers. It is operated with a commercial handheld ultrasonic testing device used in civil engineering. In order to increase the generated sound pressure of the borehole probe, the number of transducers will be increased to 32 for the novel probe. In addition, a time-controlled excitation of the individual probes is used to steer a focused sound beam to a certain angle and focus on a certain distance based on calculated time delays. Hence, the sensitive test volume is more limited, and the signal-to-noise ratio of the received signals improved. This paper presents the validation of the newly developed phased array borehole probe by beam simulation and experimentally on a half-cylindrical test specimen to investigate the directional characteristics. In combination with geophysical reconstruction method, an optimised radiation characteristic of the probe, an improved signal quality and thus an increased reliability of the results in imaging is expected. This is of great importance to construct safe sealing structures needed for radioactive or toxic waste disposal.
A novel ultrasonic borehole probe is developed for the quality assurance of sealing structures in radioactive waste repositories using existing research boreholes. The goal is to examine the sealing structures made of salt concrete for possible cracks, delamination, and embedded objects. A prototype probe uses 12 individual horizontal dry point contact (DPC) shear wave transducers separated by equidistant transmitter/receiver arrays of six individual transducers each. It is operated with a commercially available portable ultrasonic flaw detector used in the civil engineering industry. To increase the generated acoustic pressure of the borehole probe, the number of transducers in the novel probe is increased to 32. In addition, a timed excitation of each probe directs a focused beam of sound to a specific angle and distance based on calculated time delays. This limits the sensitive test volume and improves the signal-to-noise ratio of the received signals. This paper presents the validation of the newly developed phased array borehole probe by investigating the directivity through beam computation in the CIVA software and experimental investigations on a semi-cylindrical test specimen. It is expected that an optimised radiation pattern of the probe in combination with geophysical reconstruction methods will improve the signal quality and thus increase the reliability of the imaging results. This is of great importance for the construction of safe sealing structures needed for the disposal of radioactive or toxic waste.
The excellent mechanical properties of elastomer seals at a wide range of temperatures as well as their high versatility and recovery potential under several load conditions make these materials well suitable for the application in containers designed for transport and disposal of negligible heat generating radioactive waste. While a seal exchange at defined intervals is typical in many conventional applications, it is impossible or at least hard to perform when principles of minimization of radiation exposure have to be considered which prohibit an avoidable cask handling. An extensive knowledge of the change of the elastomer properties during aging and the availability of reliable end-of-lifetime criteria to guarantee the safe enclosure of the radioactive material for the required time are mandatory. As BAM is involved in most of the national cask licensing procedures and in the evaluation of cask-related long-term safety issues, great efforts have already been made and are still ongoing to scientifically support this task.
Among other representative types of elastomers, specimen made from ethylene propylene diene rubber (EPDM) were tested before, during and after aging to capture the with respect to application most important of their complex mechanical properties.
Exemplary results of these investigations were used to calibrate material models implemented in the commercial finite element software ABAQUS/Standard®. The finite element model already presented in previous works uses a sequential temperature displacement coupling. The calculated compression stress relaxation (CSR) and compression set (CS) values do satisfactorily match the experimental results.
In many investigations performed at BAM both values (CSR and CS) were identified as key indicators of elastomer’s long-term performance.
However, the possibility to correlate these equivalent indicators with performance values such as tightness and leakage rate, measurable in the mounted state, is an important goal of our future work. In the presented study the ABAQUS® feature of “pressure penetration” is introduced in the suggested finite element model for this purpose. It provides the possibility to simulate the penetration of a gas into a possible gap between flange and O-ring causing an opening of a leakage path. Three dimensional and axis-symmetric finite element models were generated to represent flat and grooved flanges of different dimensions.
The sensitivity of the feature to several input parameters is investigated and the observed behavior of the O-ring is correlated with the results of performed leakage tests.
Die verteilte faseroptische akustische Sensorik (DAS) wird vorgestellt, hinsichtlich ihrer Leistungsfähigkeit beschrieben und verschiedenste Anwendungsfelder, z.B. für das Zustandsmonitoring oder für seismische Messungen, beispielhaft gezeigt. Zudem werden mögliche Anwendungsfelder von DAS für den Kontext Kerntechnische Entsorgung (KTE) und Rückbau aufgezeigt.
Aging of elastomer O-rings and PE neutron shielding materials for radioactive waste containers
(2021)
Our institution BAM in Berlin, Germany is concerned with research and testing of materials in the context of safety in chemistry and technology. Our working group is involved in the licensing procedures of casks for radioactive waste. Besides, we’re doing research on aging and lifetime prediction of elastomer O-rings and investigate degradation and thermal expansion of PE neutron shielding materials.
Zu jedem Endlager gehört die Entwicklung eines Verschlusskonzept. Im Endlager Morsleben und an anderer Stelle werden in Abhängigkeit vom Wirtsgestein verschiedene Materialien und Bauweisen für Streckenverschlüsse entwickelt und in großem Maßstab erprobt. Hierzu gehören auch Maßnahmen der Qualitätssicherung, die Informationen über Ausführung und Zustand der Versuchsbauwerke liefern. Hierzu werden von der Bundesanstalt für Materialforschung und -prüfung (BAM) für verschiedene Forschungsvorhaben und Prüfaufträge Methoden und Geräte getestet, verbessert und vor Ort unter Tage eingesetzt. Die hier beschriebenen Verfahren lassen sich auch zu Untersuchungen im Wirtsgestein anwenden.
The EURATOM PREDIS project (http://www.predis-h2020.eu, last access:TS1) targets the development and implementation of activities for predisposal treatment of radioactive waste streams other than nuclear fuel and high-level radioactive waste. It started on 1 September 2020 with a 4 year duration. The consortium includes 47 partners from 17 member states. The overall budget of the project is EUR23.7 million, with EC contribution of EUR 14 million. The PREDIS project develops and increases the technological readiness level (TRL) of treatment and conditioning methodologies for wastes for which no adequate or industrially mature Solutions are currently available, including metallic materials, liquid organic waste and solid organic waste. The PREDIS project also develops innovations in cemented waste handling and predisposal storage by testing and evaluating.
The technical work packages align with priorities formulated within the Roadmap Theme 2 of EURAD (https://www.ejp-eurad.eu/TS2), Nugenia Global Vision (https://snetp.eu/nugenia/TS3) and with those identified by the project’s industrial end users group (EUG). The PREDIS will produce tools guiding decision making on the added value of the developed technologies and their impact on the design, safety and economics of waste management and disposal.
Four technical work packages are focusing on specific waste types: metallic, liquid organic, solid organic, and cemented wastes. For the first three, the main aim lies in processing, stabilizing, and packaging the different waste streams, e.g. by using novel geopolymers, to deliver items which are in line with national and international waste acceptance criteria. In contrast, the fourth technical work package has a different focus.
To provide better ways for a safe and effective monitoring of cemented waste packages including prediction tools to assess the future integrity development during predisposal activities, several digital tools are evaluated and improved. Safety enhancement (e.g. less exposure of testing personnel) and cost-effectiveness are part of the intended impact. The work includes but is not limited to inspection methods, such as muon imaging, wireless sensors integrated into waste packages as well as external package and facility monitoring, such as remote Fiber optic sensors. The sensors applied will go beyond radiation monitoring and include proxy parameters important for long-term integrity assessment (e.g. internal pressure). Sensors will also be made cost-effective to allow the
installation of many more sensors compared to current practice. The measured data will be used in digital Twins of the waste packages for specific simulations (geochemical, integrity) providing a prediction of future behavior.
Machine learning techniques trained by the characterization of older waste packages will help to connect the models to the current data. All data (measured and simulated) will be collected in a joint database and connected to a decision framework to be used at actual facilities.
The presentation includes detailed information about the various tools under consideration in the Monitoring of cemented waste packages, their connection and first results of the research.
Within the project SealWasteSafe, we advance construction materials and monitoring concepts of sealing structures applied for underground disposal of nuclear or toxic waste. As these engineered barriers have high demands concerning integrity, an innovative alkali-activated material (AAM) is improved and tested on various laboratory scales. This AAM has low reaction kinetics related to a preferential slow release of the heat of reaction in comparison to alternative salt concretes based on Portland cement or magnesium oxychloride cements. Hence, crack formation due to thermally induced strain is reduced. After successful laboratory scale analysis (Sturm et al., 2021), the AAM is characterised on a larger scale by manufacturing test specimens (100–300 L).
Conventional salt concrete (DBE, 2004) and the newly developed AAM are compared using two specimen geometries, i.e. cylindrical and cuboid. A comprehensive multisensor monitoring scheme is developed to compare the setting process of AAM and salt concrete for these manufactured specimens. The analysed parameters include temperature and humidity of the material, acoustic emissions, and strain variations. Passive sensor systems based on radiofrequency identification technology (RFID) embedded in the concrete, enable wireless access to temperature and humidity measurements and are compared to conventional cabled systems. Additionally, fibre-optic sensors (FOS) are embedded to record strain, but also have potential to record temperature and moisture conditions. Part of this project aims at demonstrating the high reliability of sensors and also their resistance to highly alkaline environments and to water intrusion along cables or at sensor locations. Further technical improvements were implemented so that first results clearly show the scalability of the setting process from previous small-scale AAM experiments and particularly the high potential of the newly developed approaches.
Furthermore, ultrasonic methods are used for quality assurance to detect obstacles, potential cracks and delamination. On the one hand, both active and passive ultrasonic measurements complement the results obtained from the multisensor monitoring scheme for the produced specimens. On the other hand, the unique large aperture ultrasonic system (LAUS) provides great depth penetration (up to nearly 10 m) and can thus be applied at in situ sealing structures built as a test site in Morsleben by the Federal Company for Radioactive Waste Disposal (Bundesgesellschaft für Endlagerung, BGE) as shown by Effner et al. (2021). An optimised field lay-out identified from forward modelling studies and advanced imaging techniques applied to the measured data will further improve the obtained results. To characterise the inside of the test engineered barrier and achieve a proof-of-concept, an ultrasonic borehole probe is developed to enable phased arrays that can further improve the detection of potential cracks. Modelling results and first analysis of semispherical specimens confirmed the reliability of the directional response caused by the phased arrays of the newly constructed ultrasonic borehole probe.
Overall, the project SealWasteSafe improves the construction material, multisensor monitoring concepts and ultrasonics for quality assurance. This will help to develop safe sealing structures for nuclear waste disposal. The outcomes are particularly valuable for salt as a host rock but partly also transferrable to alternative conditions.
The BAM research project BRUZL (Fracture mechanical analysis of spent fuel claddings during long-term dry interim storage) evaluates results of quasi-static Ring Compression Tests (RCT). Due to its simple realisation, the RCT is an established method for characterising the mechanical behaviour of fuel cladding, where a small tube sample is subjected to a compressive load and the relation between load and displacement is obtained. During tests at room temperature, unirradiated ZIRLO® samples with a distinct extend of radial hydrides have shown high susceptibility to sudden, brittle failure at relatively small loads. Hydride morphology and radial hydride metrics have been an integral part of the characterisation of degradation mechanisms of cladding. A software tool for the analysation of metallographic micrographs was developed. The program allows the automated calculation of hydride lengths, detection of continuous hydride structures, their statistical evaluation and visualisation. Numerical models have been established to reproduce experimental RCT results featuring an elastic-plastic material model, which was derived from inverse finite element analyses of as-received RCT samples. A cohesive zone model (CZM) is introduced to the model to reproduce load drops in the linear regime of the load vs. displacement curve. Using a traction-separation approach for the CZM, the cohesive parameters are fitted to match numerical and experimental results yielding the fracture mechanical material properties. Combining the three key aspects hydride morphology, fracture mechanics and RCT stress-state a failure criterion can be established as a result of the BRUZL project.
The project is funded by the Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561.
The topic of the presentation are distributed fiber optic measurement techniques that can be used for long-term monitoring of strain, temperature and radiation distribution in the radiation environment. The fiber optic sensors applied to the surface of waste packages provide detection and localization of cracks.
Abgeleitet aus den in Teil 3 der „Systemanalyse Konrad“ [1] definierten Lastannahmen für den „untertägigen Anlagenbereich“ müssen Verpackungen, die der Abfallbehälterklasse I (ABK I) mit dem zusätzlichen Merkmal der störfallfesten (sf) Verpackung oder der ABK II zugeordnet werden, gemäß den „Endlagerungsbedingungen, Endlager Konrad“ [2], nach einem „…Fall aus 5 m Höhe auf eine unnachgiebige Unterlage…“ bestimmten Anforderungen genügen.
Eine nähere Beschreibung der Prüfeinrichtung sowie die Definition detaillierter Anforderungen an das Aufprallfundament erfolgen in der „Produktkontrolle, Endlager Konrad“ [3]: „Die Prüfeinrichtung, bestehend aus einer Hebeeinrichtung und einem Fundament ist so zu gestalten, dass der Fall des Prüfmusters in jeder beliebigen Aufprallorientierung aus 5 m Höhe, gemessen von der Unterkante des Prüfmusters zum Fundament, möglich ist. Das Fundament ist aus Beton der Mindestgüte B 35 herzustellen und muss so beschaffen und verankert sein, dass keine Zerstörung in zahlreiche kleinere Bruchstücke erfolgt und keine horizontale Verschiebung des Fundamentes oder seiner Teile auftritt. Maße und Masse des Fundaments müssen so beschaffen sein, dass der Einfluss des Baugrundes unter dem Fundament auf den Aufprallwiderstand vernachlässigbar ist.“
Im Rahmen der Bauartprüfung für Behälter zur Endlagerung radioaktiver Abfälle mit zu vernachlässigender Wärmeentwicklung im Endlager Konrad wird die Bundesanstalt für Materialforschung und -prüfung (BAM) von der Bundesgesellschaft für Endlagerung (BGE) als Sachverständige hinzugezogen und ist in der Regel auch mit der Durchführung der notwendigen Fallprüfungen auf dem Testgelände Technische Sicherheit der BAM (BAM TTS) einschließlich der Bereitstellung der benötigten Prüfeinrichtungen sowie des regelkonformen Aufprallfundamentes beauftragt. Der Prozess, beginnend von der Gewährleistung einer von allen Beteiligten als gültig bewerteten Fallprüfung bis hin zur abschließenden Bewertung und Interpretation der Versuchsergebnisse erfordert dabei u.a. eine sorgfältige Planung und Fertigung der für das Endlager Konrad repräsentativen Fundamentplatte und deren Ankopplung an einen Untergrund im Sinne der „Regulations for the Safe Transport of Radioactive Material“ [4], [5] wie er zum Beispiel auf den beiden Fallversuchsanlagen des BAM Testgelände Technische Sicherheit (BAM TTS) zur Verfügung steht.
Um diese Anforderungen zu erfüllen muss zum einen der Erhalt der Integrität der Fundamentplatte beim Fallversuch durch eine geeignete Auswahl und kraftflussgerechte Anordnung der Stabstahl-Bewehrung sowie deren Verankerung garantiert und damit eine unzulässige Energieaufnahme durch Risse sowie eine unzulässige „… Zerstörung (der Fundamentplatte) in zahlreiche kleinere Bruchstücke …“, wie in [3] gefordert, verhindert werden. Außerdem ist zu gewährleisten, dass der Beton die Anforderungen an die vorgeschriebene Mindestgüte bzw. -festigkeit zum Zeitpunkt der Fallprüfung erfüllt, die Betonfestigkeit jedoch zur Vermeidung unnötig verschärfter Prüfrandbedingungen eine zu definierende Obergrenze nicht überschreitet.
Ausgehend von dem 2009 im Rahmen der KONTEC veröffentlichten Wissensstand [6] präsentiert die BAM im vorliegenden Beitrag die wesentlichen Ergebnisse der zwischenzeitlich bei der Planung und Durchführung von zahlreichen Fallprüfungen gewonnenen Erfahrungen bzgl. Spezifikation und Fertigung der für die Fallversuche notwendigen Fundamentplatten sowie deren Anbindung an den Untergrund. Neben den immer wieder notwendigen Anstrengungen zur Gewährleistung einer im Rahmen der Definition korrekten Betondruckfestigkeit am Tag der Fallprüfung wird die Anpassung der Bewehrungsführung an verschiedene Fallpositionen diskutiert.
Zur Ergänzung des gültigen Regelwerks entsteht parallel zu diesem Dokument eine Fachnotiz, in der BGE und BAM die Vorgaben und Spezifikationen zur regelwerkskonformen Herstellung und Montage von Fundamentplatten für Fallprüfungen im Rahmen der Behälterbauartprüfungsverfahren für das Endlager Konrad zusammenfassen und veröffentlichen werden. Diese Fachnotiz soll detaillierte Vorgaben und Spezifikationen zur Planung, Beauftragung, Herstellung und abschließenden Prüfung eines solchen Fundamentes und dessen Anbindung an den Untergrund enthalten. Damit soll allen beteiligten Organisationen wie Antragstellern, Sachverständigen und BGE ein langfristig verlässlicher Leitfaden im Hinblick auf die Durchführung anforderungsgerechter Fallprüfungen unter definierten und reproduzierbaren Randbedingungen im Rahmen der Bauartprüfungsverfahren für das Endlager Konrad an die Hand gegeben werden.
With integration of the new para 613A into SSR-6 [1] the consideration of ageing mechanisms is now obligatory for the design of packages and their approval. In addition, para 809(f) of SSR-6 [1] requires consideration of the effects of storage on ageing mechanisms, safety analyses and operation and maintenance instructions.
German competent authorities Bundesanstalt für Materialforschung und -prüfung (BAM) and Bundesamt für die Sicherheit der nuklearen Entsorgung (BASE) are considering the aspect of ageing in approval procedures. Ageing assessment is mainly focused on dual purpose casks (DPC) package designs which are long-term stored in interim storage facilities.
Transport containers with radioactive material are usually shipped in Germany with wooden-filled impact limiters. Alternative energy-absorbing materials for this purpose can be polyurethane foam, for example. In order to adequately determine the characteristics for a selected PU foam, various experimental investigations must be carried out. Classically, PU foam specimens are tested under uniaxial compression, confined compression and a tensile test. Using the experimental data and selected material models from the literature, the parameters can be determined to describe and apply material models.
Transport containers with radioactive material are usually shipped in Germany with wooden-filled impact limiters. Alternative energy-absorbing materials for this purpose can be polyurethane foam, for example. In order to perform adequate simulations with PU foam, experimental and numerical investigations must be carried out. With the help of a series of experimental test runs, a material model is developed, tested and compared in LS-Dyna.
The containment system of transport packages for spent nuclear fuel and high-level waste usually includes bolted lids with metal gaskets. The specified transport condition imply high loading on the lids and the bolt connections of the package. The response of the lid systems on these load conditions is generally investigated by drop tests or numerically. BAM has started a research project to get a better understanding about the behavior of prestressed bolt connection under loadings typical for drop tests.
The presentation gives an overview over three current fire testing activities for testing transport packages for the transport of high-level radioactive material. Packages for the transport of high-level radioactive material must withstand severe hypothetical accidents. Regulatory test conditions shall cover these severe accident conditions and consist of mechanical tests and a following thermal test. To withstand the mechanical tests heavy weight packages are often designed with impact limiters consisting of wood encapsulated in steel sheets. The thermal test is defined precisely in the IAEA-regulations as a 30 minute fully engulfing 800 °C fire. After the fire phase a pre-damaged impact limiter might continue burning or smoldering and influence the cask thermal behavior with its energy release. The energy transferred from the impact limiter to the cask is of importance for the safety of transport packages. A full-scale fire test with an impact limiter of 2.3 m in diameter and filled with spruce wood was designed and performed. The impact limiter continued burning for 3 days. Energy transfer and temperature measurements were performed. A new test is designed to examine pressure build up and possible mechanical failures in an undamaged impact limiter with a diameter of 1.9 m. The test is designed to measure heat flux from the impact limiter in case of its ignition and burning. To furthermore examine the burning behavior of steel encapsulated wood piles, tests are prepared with wood fire containers and an infrared ignition source. A fire reference package for calorimetric tests was designed. The fire reference package design consists of a closed steel sheet cylinder with a length of 1,500 mm, an exterior diameter of 1,050 mm, and a wall thickness of 10 mm. The cylinder was filled with refractory insulation material and instrumented with thermocouples distributed all over the cylinder. The measured local steel sheet temperatures allow the determination of local as well as overall integral heat fluxes versus time and versus surface temperature. Currently, tests are planned with a new and relatively small fire reference package design. The new design will have a steel sheet cylinder with a length of 182 mm, an exterior diameter of 120 mm. Additionally heat flux sensors will be used in the new fire reference package test design. The propane gas fire will be adjusted with respect to the outcome of the fire tests to meet the requirements of the IAEA-fire.
With integration of the new para 613A into SSR-6 [1] the consideration of ageing mechanisms is now obligatory for the design of packages and their approval. In addition, para 809(f) of SSR-6 [1] requires consideration of the effects of storage on ageing mechanisms, safety analyses and operation and maintenance instructions.
German competent authorities Bundesanstalt für Materialforschung und -prüfung (BAM) and Bundesamt für die Sicherheit der nuklearen Entsorgung (BASE) are considering the aspect of ageing in approval procedures. Ageing assessment is mainly focused on dual purpose casks (DPC) package designs which are long-term stored in interim storage facilities. For these package designs, the evaluation of ageing management is now mandatory for the maintenance of the package design approvals with a validity period of 5 years and beyond. The ageing management includes amongst others a gap analysis, the assessment of ageing effects and operational experiences during operation and interim storage.
BAM works on the compilation of a guideline for implementation of paras 613A, 809(f) and 809(k) for packages requiring competent authority approval at the application procedure in Germany. The paper describes essential items of ageing mechanisms and will give a foresight to the ageing management evaluation by BAM.
Bei der Auswahl eines Standortes für ein Endlager für hochradioaktive Abfälle (Wärme entwickelnde radioaktive Abfälle und ausgediente Brennelemente) in Deutschland werden auf Grundlage des Standortauswahlgesetzes von 2017 die drei potenziellen Wirtsgesteine Steinsalz, Tonstein und Kristallingestein in Betracht gezogen. Die Auswahl eines konkreten Endlagerstandortes steht dabei in engem Zusammenhang mit den für die jeweiligen Wirtsgesteine geeignet erscheinender Endlagersystemen. Zentraler Bestandteil dieser Endlagersysteme sind die Endlagerbehälter, deren Gestaltung und Auslegung maßgeblich die Randbedingungen für die Transport- und Einlagerungstechnik vorgibt und zudem die Nachweise zur Betriebs- und Langzeitsicherheit entscheidend prägt. Daher ist es für die Behälterentwicklung erforderlich, die an Endlagerbehälter zu stellenden Anforderungen systematisch herzuleiten. Auf dieser Grundlage können anforderungsgerechte Endlagerbehälter zielgerichtet für die jeweils zu untersuchenden Endlagerstandorte in unterschiedlichen Wirtsgesteinen entwickelt werden.
Dieser Beitrag fasst die wesentlichen Ergebnisse des von BAM und BGE TECHNOLOGY GmbH im Zeitraum 2017 bis 2020 bearbeiteten und vom Bundesministerium für Wirtschaft und Energie (BMWi) geförderten Forschungsvorhabens KoBrA zusammen, die in einem Abschlussbericht und vier zugehörigen Teilberichten umfassend dokumentiert sind.
During an extensive test programme at the Bundesanstalt für Materialforschung und prüfung, material property changes of EPDM O-rings were investigated at different ageing times and an ageing temperature of 125 °C. To exclude possible diffusion limited oxidation (DLO) effects that can distort the data, IRHD micro hardness measurements were performed over the cross-section of compressed O-rings. Continuous stress relaxation measurements were conducted on samples free of DLO effects. The additional effect of physical processes to irreversible chemical ones during a long-term thermal exposure is quantified by the analysis of compression set measurements under various test conditions. By combining the different experimental methods, characteristic times relative to the degradation processes were determined. On the basis of experimental data, a microphysically motivated model that takes into account reversible and irreversible processes was developed. The parameter identification strategy of the material model is based on our experimental investigations on homogeneously aged elastomer O-rings. The simulated results are in good agreement with the experiments.
The European Joint Programme on Radioactive Waste Management (EURAD) deals in work package (WP) 8 with Spent Fuel Characterization (SFC). Inspired by the EURAD activities, an international Coordinated Research Project (CRP) on SFC was established by the International Atomic Energy Agency (IAEA). The EURAD WP SFC participants are collaborating as a team on the IAEA CRP on SFC. The EURAD WP SFC project is divided into 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. During an IAEA consultancy meeting, BAM as leader of Task 3 reported on the current status of the research work.
The main task of the division "Safety of Storage Containers" at BAM is the safety assessment of storage containers for radioactive waste. A sufficient sealing of bolted lid systems for the safe containment of the waste for transport, storage and disposal is very important. Extensive knowledge of the change of the elastomer’s properties during ageing and the availability of reliable end-of-lifetime criteria is mandatory to guarantee long term safe enclosure.
In a long-term test programme over 5 years we have studied the degradation and the change of sealing properties of several elastomers, including EPDM, at four different ageing temperatures (75 °C, 100 °C, 125 °C and 150 °C). Compression stress relaxation (CSR) and compression set (CS) experiments were carried out. It has been found that when the data does not cover a sufficient time frame necessary for the evolution of the degradation of a chosen property, a curvature in the Arrhenius relationship is observed. For CSR, the curvature was observed for samples aged up to 186 days. As for CS experiments, the curvature was detected for sample aged up to 2 years. To cover the possible lack of experimental long-term data at low temperatures, a numerical model for CSR was developed and longer ageing times for the simulation were adopted. A degradation-rate based model for the evolution of degradative processes is proposed. The main advantage of the model is the possibility to quickly validate the interpolation at lower temperatures within the range of slower chemical processes without forcing an Arrhenius straight-line extrapolation. The model was also applied to CS experiments and validated by the 5 years experimental results where the curvature was gone, and the degradation property has followed an Arrhenius relationship. In the figure the contribution of the two degradative processes are shown over CS and CSR respectively.
Long term monitoring of concrete structures using innovative NDT and SHM approaches: NDT-CE 4.0
(2021)
In Germany, a large amount of reinforced and/or prestressed concrete structures are approaching the end of their life cycle. Many of them already show a significant reduction of operability or capacity. The issue is most obvious, but not limited to traffic or energy related infrastructure. These structures have to be inspected and assessed on a regular basis. More and more permanent monitoring systems are installed to follow the degradation and to give pre-warning before failure.
While the basic mandatory inspection (e. g. of bridges) is still limited to visual examination and tap tests, more advanced investigation tools as ultrasound or radar are used more and more often. Monitoring e. g. by acoustic emission systems is increasingly used e.g. for detection of wire breaks in prestressed structures. The presentation gives an overview on the state of the art and an outlook on innovative systems, including the integration in modern data processing, evaluation, display and archive systems.
To provide better means for a safe and effective monitoring of cemented waste packages including prediction tools to assess the future integrity development during pre-disposal activities, several digital tools are evaluated and improved in the frame of the EC funded project PREDIS. Safety enhancement (e. g. less exposure of testing personnel) and cost effectiveness are part of the intended impact.
The work includes but is not limited to inspection methods such as muon imaging, wireless sensors integrated into waste packages as well as external package and facility monitoring such as remote fiber optical sensors. The sensors applied will go beyond radiation monitoring and include Proxy parameters important for long term integrity assessment (e. g. internal pressure). Sensors will also be made cost effective to allow the installation of much more sensors compared to current practice.
The measured data will be used in digital twins of the packages for specific simulations (geochemical, integrity) providing a prediction of future behavior. Machine Learning techniques trained by the characterization of older packages will help to connect the models to the actual data.
All data (measured and simulated) will be collected in a joint data base and connected to a decision framework to be used at actual facilities.
The presentation includes detailed information about the various tools under consideration, their connection and first results of our research.
The workpackage 7 of the EC funded project PREDIS deals with cemented low and intermediate level radioactive waste. The focus is on NDT/monitoring technologies, digiutal twins and data/decision frameworks to optimize facilitay and waste package management. The talk describes the project setup and planned activities.
Elastomer seals are used in many industrial applications due to their excellent mechanical properties at a wide range of temperatures. Their high versatility and recovery potential under several load conditions make them well suitable for the application in containers designed for transport, storage and/or disposal of radioactive materials. In containers for low and intermediate level radioactive waste, elastomer seals are used as barrier seals, and as auxiliary seals in storage and transportation casks (dual purpose casks) for heat generating radioactive waste, such as spent fuel and high-level waste. While a seal exchange at defined intervals is typical in many conventional applications, it is impossible or at least hard to perform when principles of minimization of radiation exposure have to be considered and prohibit an unnecessary cask handling. An extensive knowledge of the change of the elastomer’s properties during aging and the availability of reliable end-of-lifetime criteria to guarantee the permanent safe enclosure of the radioactive material is mandatory. As BAM is involved in most of the national cask licensing procedures and in the evaluation of cask-related long-term safety issues, great efforts have been already made and are still planned to scientifically support this task.
Compression stress relaxation and compression set were identified as key indicators of elastomer long-term performance and quantitatively investigated in comprehensive test programs. Among other representative types of elastomers, specimens made from ethylene propylene diene rubber (EPDM) were tested before, during and after aging to capture the most important of their complex mechanical properties.
In the presented study, exemplary results were used to simulate the compression stress relaxation and the compression set of elastomer O-rings during aging. Regarding the influence of temperature, the time temperature superposition principle is applied in the relaxation analysis of elastomer O-rings. The proposed model is implemented in the commercial finite element software ABAQUS/Standard® [1] with a sequential temperature displacement coupling. Numerical results match the experimental compression stress relaxation measurements well. The prediction of compression set values after long-term aging shows a relatively good agreement with the experimental results. Nevertheless, all input parameters derived from the specimen tests, additional assumptions concerning boundary conditions and modeling strategy are discussed with regard to the identified slight discrepancies.
The possibility to extend the finite element model to represent the O-ring seal’s ability to recover after a (fast) partial release is taken into account
Faseroptische Strahlungssensoren ermöglichen ein räumlich verteiltes Online-Monitoring entlang der gesamten Sensorfaser. Verschiedene Messverfahren nutzen überwiegend zwei durch die ionisierende Strahlung in optischen Fasern induzierte Effekte, d. h. die Erhöhung der optischen Dämpfung und die damit verbundene Veränderung des Brechungsindex des Fasermaterials. Die Sensitivität der Strahlungssensoren lässt sich je nach Anwendung durch die Wahl der Dotierstoffe im Glasfaserkern sowie der Art des Polymermaterials des POF-Sensors beeinflussen. Ferner ist die Sensitivität auch durch den Betrieb bei geeigneten Wellenlängen einstellbar. Bei dem sensorischen Einsatz von faseroptischen Sensoren muss auch die vorhandene Ausheilung des Sensors sowie die Abhängigkeit der erfassten Messsignale von der Temperatur und der Dosisleistung berücksichtigt werden.
Packages for the transport of spent nuclear fuel shall meet the International Atomic Energy Agency regulations to ensure safety under different Transport conditions. The physical state of spent fuel and the fuel rod cladding as well as the geometric configuration of fuel assemblies are important inputs for the evaluation of package capabilities under these conditions. Generally, the mechanical behavior of high burn-up spent fuel assemblies under Transport conditions shall be analyzed with regard to the assumptions which are used in the containment and criticality safety analysis. In view of the complexity of the interactions between the fuel rods as well as between the fuel assemblies, basket, and cask containment, the exact mechanical analysis of such phenomena is nearly impossible. The gaps in information concerning the material properties of cladding and pellet behavior, especially for the high burn-up fuel, make the analysis more complicated additionally. As a result, enveloping analytical approaches are usually used by BAM within the safety assessment of packages approved for transport of spent nuclear fuel.
To justify the safety margins of such approaches additional analyses are necessary. In this paper, numerical simulations of a segment of a spent fuel assembly are presented. The segment modeled represents the part of a generalized BWR fuel assembly between two spacers. Explicit dynamic finite element calculations are performed to simulate the spent fuel behavior under regulatory defined accident conditions of transport. A beam element formulation is used for the modeling of the fuel rods representing the compound consisting of claddings and fuel pellets. The load applied is gathered from experimental drop tests with spent fuel casks performed at BAM. A hot cell bending test performed at JRC Karlsruhe is the basis for obtaining the material behavior of the fuel rods. The material properties are determined by
simulating the test setup of JRC and optimizing the results to fit the
experimental load deflection curve. The simulations of the fuel Assembly segment are used to get a better understanding about the loads on fuel rods under accident conditions of transport.
An established method for characterizing cladding material is the Ring Compression Test (RCT), where a small, cylindrical sample of the cladding tube is subjected to a compressive load. This test is a laboratory representation of a fuel rod load case and has shown a high susceptibility to failure under the occurrence of radial hydrides. Certain hydride morphologies and low temperature might even lead to brittle fracture at very small loads. The BAM research project BRUZL (Fracture mechanical analysis of spent fuel claddings during long-term dry interim storage) aims to study the results of quasi-static RCTs and to establish numerical models to gain an in-depth understanding of the stress state during such experiments. Using a fracture mechanical approach, cases of sudden failure during the RCT procedure are used to characterize the material behaviour and establish a failure criterion.
As an integral part of the project BRUZL, quasi-static Ring Compression Tests have been performed to identify all experimental details, which might be helpful for numerical modelling. Unirradiated samples of the cladding material ZIRLO® have been subjected to hydrogen charging and a thermo-mechanical treatment for radial hydride reorientation. Sample preparation, testing procedures, and analysis results are presented. A numerical model has been established and an elastic-plastic material model was derived from as-received RCT samples by inverse finite element analyses. Cohesive zone modelling has been implemented to reproduce sudden load drops during RCT.
The project is funded by the Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561.
Understanding the mechanical behaviour of fuel cladding is decisive for an extended dry interim storage period beyond the currently licensed period of 40 years in Germany and subsequent handling, transport, and disposal considering normal and accidental conditions. Since fuel cladding acts as the first barrier for spent fuel, its integrity should be demonstrated to enable safe and secure discharge and handling options after interim storage and subsequent transportation in preparation of final disposal. An established method for characterizing cladding material is the Ring Compression Test (RCT), where a small, cylindrical sample of the cladding tube is subjected to a compressive load. This test is a laboratory representation of a fuel rod load case and has shown a high susceptibility to failure under the occurrence of radial hydrides. Certain hydride morphologies and low temperature might even lead to brittle fracture at very small loads. The BAM research project BRUZL (Fracture mechanical analysis of spent fuel claddings during long-term dry interim storage) aims to study the results of quasi-static RCTs and to establish numerical models to gain an in-depth understanding of the stress state during the experiments. Using a fracture mechanical approach, cases of sudden failure during the RCT procedure are used to characterize the material behaviour and establish a failure criterion. As an integral part of the project BRUZL, quasi-static Ring Compression Tests have been performed to identify all experimental details, which might be helpful for numerical modelling. Unirradiated samples of the cladding material ZIRLO® have been subjected to hydrogen charging and a thermo-mechanical treatment for radial hydride reorientation. Sample preparation, testing procedures, and analysis results are presented. A numerical model has been established and an elastic-plastic material model was derived from as-received RCT samples by inverse finite element analyses. Cohesive zone modelling has been implemented to reproduce sudden load drops during RCT. The project is funded by the Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561.
Outcomes of Three Large-Scale Fire Reference Tests Conducted in Propane Gas Fire Test Facility
(2020)
Packages for the transport of high-level radioactive material are designed to withstand severe accidents. Hypothetical severe accident conditions are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials. One of these accident conditions is the thermal test, mainly consisting of a 30 minute fully engulfing 800°C pool fire or an equally severe fire test. The heat fluxes into the package depend substantially on the fire characteristics and the surface temperature of the package. Fire tests can be performed at BAM on a propane gas fire test facility. In order to investigate the heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for multiple use. The package represented the outer geometry of a specific transport cask for radioactive waste. The fire reference package is a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 1500 mm and a diameter of 1050 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material. Three open-air fire tests were performed in the BAM propane gas fire test facility. The flames exposure time period varied slightly for the fire tests. The wind direction as well as the wind speed were measured and changed between and during the tests. Test stand parameters such as wind shield location and propane gas volume flow were chosen constant for the three tests. The locally measured fire reference package steel sheet temperatures were used for the calculation of heat fluxes as function of time and surface temperature. The measured temperatures allowed further calculations. In a first approach effective fire characteristics of the propane gas fire, including the flame temperature, the fire convection coefficient and a radiation exchange coefficient mathematically describing the determined average heat flux over the surface temperature were calculated.
Prediction of dynamic effects of reinforced concrete structures under impact loading is a technical challenge. This is a consequence of the great variability of the physical properties resulting from the wide adaptability of reinforced concrete and a consequence of the wide range of impact loading. Experiments and numerical investigations are normally used on a small scale to address the problem. In this paper, impact tests on reinforced conrete plates with the lateral dimensions of 1.5 m x 1.5 m and a thickness of 30 cm are presented. In bending reinforcement, besides the velocity two properties are varied, the diameter and the spatial distribution of the rebars. Experiments are performed at the Otto-Mohr-Laboratory of the Institute of Concrete Structures of the Technische Universit¨at Dresden. Due to the accelerated fall of the impactor the velocity ranges between 20 and 70 m/s. In addition to the measured quantities such as bearing forces, accelerations are also measured at 4 different positions on and under the plate, as well as the deflection at several positions. The measured data are used for the analysis of the damage form and the numerical examinations with the program Ansys Autodyn and the material model after Drucker-Prager. Numerical investigations support the tests, with detailed analysis of individual effects. These numerical computations and the planar tomographic investigations were carried out at BAM in Berlin. With the help of planar tomographic evaluation, the damaged structure is made visible and compared with the numerical results. Influences of the bending reinforcement are explained on the basis of damage evaluation in the local area and on selected measured values. In addition to the test evaluation, the tomographic and numerical methods are presented.
In Germany, spent nuclear fuel and vitrified high active waste is stored in dual purpose casks (DPC) at interim storage facilities. In order to ensure the transportability of the DPC to a final repository in future, the maintenance of the package design approval is realized. Therefore, the assessment of possible ageing effects during interim storage is necessary to ensure an evaluation of the transportability. The lecture presents BAM's current ageing evaluation concept in the field of trasnport law.
The containment system of transport packages for spent nuclear fuel and high-level waste usually includes bolted lids with metal gaskets. The specified transport condition imply high loading on the lids and the bolt connections of the package. The response of the lid systems on these load conditions is generally investigated by drop tests or numerically. BAM has started a research project to get a better understanding about the behavior of prestressed bolt connection under loadings typical for drop tests.
Packages for the transport of high-level radioactive material are designed to withstand severe accidents. Hypothetical severe accident conditions are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials. One of these accident conditions is the thermal test, mainly consisting of a 30 minute fully engulfing 800°C pool fire or an equally severe fire test. The heat fluxes into the package depend substantially on the fire characteristics and the surface temperature of the package. Fire tests can be performed at BAM on a propane gas fire test facility. In order to investigate the heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for multiple use. The package represented the outer geometry of a specific transport cask for radioactive waste. The fire reference package is a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 1500 mm and a diameter of 1050 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material. Three open-air fire tests were performed in the BAM propane gas fire test facility. The flames exposure time period varied slightly for the fire tests. The wind direction as well as the wind speed were measured and changed between and during the tests. Test stand parameters such as wind shield location and propane gas volume flow were chosen constant for the three tests. The locally measured fire reference package steel sheet temperatures were used for the calculation of heat fluxes as function of time and surface temperature. The measured temperatures allowed further calculations. In a first approach effective fire characteristics of the propane gas fire, including the flame temperature, the fire convection coefficient and a radiation exchange coefficient mathematically describing the determined average heat flux over the surface temperature were calculated.