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Eingeladener Vortrag
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Storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage. Beside consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case. Not only the technical objects are subject of aging but also the safety demonstration basis is subject of "aging" due to possible changes of regulations, standards and scientific/technical knowledge. The basic document, defining the transport safety conditions, is the package design safety report (PDSR) for the transport version of the DPC. To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the PDSR, in connection with periodic renewals of package design approval certificates, have to be carried out. The main reviewing tool is a gap analysis. A gap analysis for a PDSR is the assessment of the state of technical knowledge, standards and regulations regarding safety functions of structures, systems and components.
Die Regelwerksanforderungen für Verpackungen für den Transport radioaktiver Stoffe basieren auf den Empfehlungen der IAEA und werden über die verkehrsträgerspezifischen internationalen Regelwerke in nationales Recht umgesetzt.
Um die Anforderungen des Regelwerkes zu erläutern, entwickelt die Bundesanstalt für Materialforschung und -prüfung (BAM) Gefahrgutregeln (BAM-GGR), welche die Entwickler, Hersteller und Betreiber von Behältern über die von der BAM angewandten Verfahren bei der Umsetzung gesetzlicher Bestimmungen informieren und unterstützen sollen.
In Germany the concept of dry interim storage of spent fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years. The casks being used have an approved transport package design in accordance with the international IAEA transport regulations as well. The storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage.
Besides consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case.
Not only the technical objects are subject of aging but also the safety demonstration basis is subject of “aging” due to possible changes of regulations, standards and scientific/technical knowledge.
To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the Package Design Safety Report (PDSR), in connection with periodic renewals of package design approval certificates, have to be carried out.
Packages for the transport of radioactive material are generally equipped with specific
structures (basket) to support the radioactive content in defined position. The safety function
of the basket depends on the kind of transported inventory. In case of transport cask for spent
fuel, the basket design has to ensure the subcriticality of the fissile material in all conditions
of transport in particular. Therefore the evaluation of structural integrity and neutron
absorption capability of the basket is an important part of complete safety analysis. Sufficient
heat transfer to maintain fuel assembly and cask temperature within allowable limits has to be
verified as well. Corrosion resistance is an additional requirement on basket materials owing
to contact with water during loading and unloading operations.
Computational and experimental methods or their combination along with additional material
and component tests can be used to analyse the mechanical and thermal basket behaviour
under transport conditions defined in IAEA regulations. By deciding between the analysis
methods, the design features (including material selection concept) as well as specific safety
function should be accounted.
In approval procedures of transport packages for radioactive materials, the competent
authority mechanical and thermal safety assessment is carried out in Germany by BAM. Some
questions of safety evaluation of basket designs are discussed in this paper based on the BAM
experience within approval procedures. The paper focuses primarily on the mechanical
behaviour of baskets with regard to the assumptions that have to be used in the criticality
safety demonstration. The state of the art methodologies for computational basket stress and
deformation analysis as well as for interpretation of drop tests results are presented.
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.
The first German package design approval certificate for a dual purpose cask intended for loading with damaged spent nuclear fuel was issued recently. BAM as part of the competent authority system in Germany carried out a comprehensive assessment procedure with respect to the mechanical and thermal design, the release of radioactive material and the quality assurance aspects of manufacturing and operation. Packages for the transport and storage of radioactive material have been assessed by BAM for many years, thus the common assessment procedure is well-known and good practice. Up to now only SNF without defects or HLW with well-defined properties were designated for long-term Interim storage and transports afterwards. Due to Germany’s nuclear phase out all other kinds of spent nuclear fuel in particular damaged spent nuclear fuel shall be packed as well. Damaged spent nuclear fuel needs a tight closure with Special encapsulations and clearly defined properties in Germany. In addition, these encapsulations shall be long-term durable, because they are not accessible after loading in a packaging within periodical inspections. The main difference to Standard package components is that encapsulations with a permanent closure achieve their specified conditions not after manufacturing but only during operation, after loading and closing. To ensure compliance with the specific conditions, special measures for quality assurance are necessary during operation of each encapsulation, e.g. drying and sealing, which were assessed by BAM. The present paper gives an overview of the conducted assessment from BAM and point out the findings concerning to the special closure lid of the approved encapsulation, which is screwed and welded. A wide verification concept is necessary to show the specific tightness under transport conditions. Together with quality assurance measures during first operation steps these encapsulations with damaged spent nuclear fuel can be handled like standard fuel assemblies in approved package designs.
BAM (Federal Institute for Material Research and Testing) is the competent authority for mechanical and thermal safety assessment of transport packages for spent fuel and high level waste (HLW) in Germany. In context with package design approval of the new German HLW cask CASTOR HAW28M, BAM performed several drop tests with a half-scale model of the CASTOR HAW/TB2. The test model was manufactured by GNS (Gesellschaft fur Nuklear Service mbH) and tested under accident transport conditions on the 200 tons BAM drop test facility at the BAM Test Site Technical Safety. For this comprehensive test program the test specimen CASTOR HAW/TB2 was instrumented at 21 measurement planes with altogether 23 piezo-resistive accelerometers, five temperature sensors and 131 tri-axial strain gauges in the container interior and exterior, respectively. The strains of four representative lid bolts were recorded by four uniaxial strain gauges per each bolt. Helium leakage rate measurements were performed before and after each test in the above noted testing sequence. The paper presents some experimental results of the half-scale CASTOR HAW/TB2 prototype (14,500 kg) and measurement data logging.
Federal Institute for Materials Research and Testing (BAM) is the competent authority for mechanical and thermal safety assessment of transport packages for spent fuel and high level waste in Germany. In context with package design approval of the new German high level waste cask CASTOR® HAW28M, BAM performed several drop tests with a half scale model of the CASTOR® HAW/TB2. The cask is manufactured by Gesellschaft für Nuklear Service mbH and was tested under accident transport conditions on the 200 tons BAM drop test facility at the BAM Test Site Technical Safety. For this comprehensive test program, the test specimen CASTOR® HAW/TB2 was instrumented at 21 measurement planes with altogether 23 piezo resistive accelerometers, five temperature sensors and 131 triaxial strain gauges in the container interior and exterior respectively. The strains of four representative lid bolts were recorded by four uniaxial strain gauges per each bolt. Helium leakage rate measurements were performed before and after each test in the above noted testing sequence. The paper presents some experimental results of the half scale CASTOR® HAW/TB2 prototype (14 500 kg) and measurement data logging. It illustrates the extensive instrumentation and analyses that are used by BAM for evaluating the cask performance to the mechanical tests required by regulations. Although some of the quantitative deceleration, velocity and strain values cannot be shown because of confidentially issues, they are provided qualitatively to illustrate the types of measurements and methodologies used at BAM.
Die Sicherheitsanforderungen, die an Behälter für den Transport radioaktiver Stoffe auf öffentlichen Verkehrswegen gestellt werden, orientieren sich an dem Gefährdungspotenzial des radioaktiven Inhalts. So müssen für den Transport abgebrannter Brennelemente oder hochradioaktiver Abfalle aus nuklearen Anlagen Behälter eingesetzt werden, die auch schweren Unfällen standhalten. Zu ihnen gehören z. B. die CASTOR®-Behälter (Cask for Storage and Transport of Radioactive Material).