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Transportierbarkeit von Transport- und Lagerbehältern für radioaktive Stoffe, welche nach eine längerfristigen trockenen Zwischenlagerung
transportiert werden müssen. Überblick über die aktuellen Regelwerke und Richtlinien sowie die Zeitpläne für das Finden eines Endlagers für hochradioaktive Stoffe und die damit verbundene verlängerte Zwischenlagerung. Des Weiteren werden das für die verlängerte Zwischenlagerung notwendige Alterungsmanagement und die noch erforderlichen Forschungstätigkeiten vorgestellt.
Transportierbarkeit von Transport- und Lagerbehältern für radioaktive Stoffe, welche nach eine längerfristigen trockenen Zwischenlagerung
transportiert werden müssen. Überblick über die aktuellen Regelwerke und Richtlinien sowie die Zeitpläne für das Finden eines Endlagers für hochradioaktive Stoffe und die damit verbundene verlängerte Zwischenlagerung. Des Weiteren werden das für die verlängerte Zwischenlagerung notwendige Alterungsmanagement und die noch erforderlichen Forschungstätigkeiten vorgestellt.
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.
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 the spent fuel and the fuel rod cladding as well as the geometric configuration of fuel assemblies are important inputs for the evaluation of the package capabilities under these conditions. The mechanical behavior of spent fuel assemblies shall be analyzed with regard to the assumptions which are used in the containment and criticality safety analysis. The package as a mechanical system is characterized by a complex set of interactions, e. g. between the fuel rods within the assembly as well as between the fuel assemblies, the basket, and the cask containment. This complexity makes it difficult to assume appropriate mechanical loads for the spent fuel inside the package. Another challenge is the assumption of material properties which represent the variation of the fuel rods regarding cladding material, burn-up and the operation history sufficiently. The objective of this paper is to give an overview about the current approaches and research for the evaluation of spent fuel behavior within the package design approval procedure. In this context, analytical, numerical and experimental results are discussed. The presented work is achieved within a collaboration of BAM with national and international partners.
Experimental and numerical analyses of spent nuclear fuel behavior under transport conditions
(2019)
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 the spent fuel and the fuel rod cladding as well as the geometric configuration of fuel assemblies are important inputs for the evaluation of the package capabilities under these conditions. The mechanical behavior of spent fuel assemblies shall be analyzed with regard to the assumptions which are used in the containment and criticality safety analysis. The package as a mechanical system is characterized by a complex set of interactions, e. g. between the fuel rods within the assembly as well as between the fuel assemblies, the basket, and the cask containment. This complexity makes it difficult to assume appropriate mechanical loads for the spent fuel inside the package. Another challenge is the assumption of material properties which represent the variation of the fuel rods regarding cladding material, burn-up and the operation history sufficiently. The objective of this paper is to give an overview about the current approaches and research for the evaluation of spent fuel behavior within the package design approval procedure. In this context, analytical, numerical and experimental results are discussed. The presented work is achieved within a collaboration of BAM with national and international partners.
For disposal of the research reactor of the Technical University Munich FRM II a new transport and storage cask design was under approval assessment by the German authorities on the basis of International Atomic Energy Agency (IAEA) requirements. The cask body is made of ductile cast iron and closed by two bolted lid systems with metal seals. The material of the lids is stainless steel. On each end of the cask the wood-filled impact limiters are installed to reduce impact loads to the cask under drop test conditions. In the cavity of the cask a basket for five spent fuel elements is arranged.
This design has been assessed by the Bundesanstalt für Materialforschung und -prüfung (BAM) in view to the mechanical and thermal safety analyses, the activity release approaches, and subjects of quality assurance and surveillance for manufacturing and operation of the package.
For the mechanical safety analyses of the package a combination of experimental testing and analytical/numerical calculations were applied. In total, four drop tests were carried out at the BAM large drop test facility. Two tests were carried out as a full IAEA drop test sequence consisting of a 9m drop test onto an unyielding target and a 1m puncture bar drop test. The other two drop tests were performed as single 9m drop tests and completed by additional analyses for considering the effects of an IAEA drop test sequence.
The main objectives of the drop tests were the investigation of the integrity of the package and its safety against release of radioactive material as well as the test of the fastening system of the impact limiters. Furthermore, the acceleration and strain signals measured during the tests were used for the verification of finite-element (FE) models applied in the safety analysis of the package design.
Accident safe packages for the transport of spent nuclear fuel and high-level waste shall fulfil international IAEA safety requirements. Compliance is shown by consecutive mechanical and thermal testing. Additional numerical analysis are usually part of the safety evaluation. For damage protection some package designs are equipped with wood filled impact limiters encapsulated by steel sheets.
The safety of these packages is established in compliance with IAEA regulations. Cumulative mechanical and fire tests are conducted to achieve safety standards and to prevent loss of containment. Mechanical reliability is proven by drop tests. Drop testing might cause significant damage of the impact limiter steel sheets and might enable sufficient oxygen supply to the impact limiter during the fire test to ignite the wood filling. The boundary conditions of the fire test are precisely described in the IAEA regulatory. During the test the impact limiter will be subjected to a 30 minutes enduring fire phase. Subsequent to the fire phase any burning of the specimen has to extinguish naturally and no artificial cooling is allowed.
At BAM a large-scale fire test with a real size impact limiter and a wood volume of about 3m³ was conducted to investigate the burning behaviour of wood filled impact limiters in steel sheet encapsulation. Until today burning of such impact limiters is not sufficiently considered in transport package design and more investigation is necessary to explore the consequences of the impacting fire. The objective of the large scale test was to find out whether a self-sustaining smouldering or even a flaming fire inside the impact limiter was initiated and what impact on the cask is resulting.
The amount of energy transferred from the impact limiter into the cask is of particular importance for the safety of heavy weight packages. With the intention of heat flux quantification, a new approach was made and a test bench was designed.
Im Testgelände Technische Sicherheit (TTS) verfügt die BAM über eine umfangreiche prüftechnische Infrastruktur für die mechanische und thermische Behälterbauartprüfung unterschiedlicher Behälterdesigns und -komponenten. Berichtet wird über Erprobungsprogramme und prüfmethodische Untersuchungen mit originalgroßen Prüfmustern von Transportverpackungen und Lagercontainer sowie Komponentenuntersuchungen unter Impaktbelastung. Im Großkomponentenversuch werden mit einem Prüfmuster eines Transportbehälters charakteristische Unfallbeförderungsbedingungen simuliert, dabei die Verformungen der Stoßdämpferkomponenten sowie die resultierenden maximalen Starrkörper-Verzögerungen ermittelt und analysiert. Ferner wird über das mechanische und thermische Beanspruchungsverhalten einer Umverpackung für den Transport von UF6- Zylindern im Ergebnis umfangreicher Fallsequenzen sowie einer Brandprüfung berichtet. Im Rahmen der Behälterentwicklung für den Standort Sellafield wurden mit einem dickwandigen Container für radioaktive Abfälle Fallprüfungen auf das unnachgiebige Fundament und eine Brandprüfung durchgeführt. Detaillierte Belastungsanalysen erfolgten an stoßdämpfenden Bauteilen und Materialien im Rahmen der angewandten Sicherheitsforschung.
Methoden der Bauartprüfung und –zulassung von Behältern für radioaktive Stoffe, Sicherheitsnachweis von RAM –Behältern;
Versuchstechnische Anlagen der BAM für Gefahrgutverpackungen im Originalmaßstab; Durchführung experimenteller Untersuchungen mit Transport- und Lagerbehältern für radioaktive Stoffe; Fallprüfungen mit Prüfmustern und Modellbehältern; Beanspruchungsanalyse von Behälterkomponenten und Sicherheitsbewertung der Behälterkonstruktion; Vorstellung der BAM und des Testgeländes Technische Sicherheit