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Im Zuge des Rückbaus kerntechnischer Anlagen in Deutschland müssen u. a. Großkomponenten (Dampferzeuger, Reaktordruckbehälter) über öffentliche Verkehrswege in Zwischenlager transportiert werden, in denen sie zerlegt bzw. zwischengelagert werden. Für diese Versandstücke ist eine mechanische Integritätsbewertung, nach den Vorschriften der Internationalen Atomenergie-Organisation (IAEO), notwendig, die in der Regel unter dem Einsatz einer komplexen numerischen Berechnung erfolgt.
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In the course of decommissioning of power plants in Germany large nuclear components (steam generator, reactor pressure vessel) must be transported over public traffic routes to interim storage facilities, where they are dismantled or stored temporarily. For these packages a safety evaluation considering the International Atomic Energy Agency (IAEA) transport regulations is necessary, preferably by a complex numerical analysis.
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.
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.
Im Gegensatz zu den zulassungspflichtigen Versandstücken für die gemäß der gefahrgutrechtlichen Regelwerksanforderungen eine behördlich ausgestellte Zulassung erforderlich ist, ist für die prüfpflichtigen Versandstücke lediglich eine behördliche Anerkennung und Überwachung des Managementsystems für die Auslegung, Herstellung, Prüfung, Dokumentation, den Gebrauch, die Wartung und Inspektion erforderlich.
In der Bundesrepublik Deutschland ist gemäß den Festlegungen in der Gefahrgutverordnung Straße, Eisenbahn und Binnenschifffahrt die Bundesanstalt für Materialforschung und -prüfung (BAM) zuständig für die Anerkennung und Überwachung von Managementsystemen.
Assessment of quality management for transport packages not requiring authority design approval
(2018)
The majority of transports of radioactive materials are carried out in packages which don’t need a package design approval of a competent authority. Low active radioactive materials are transported in such kind of packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well.
In Germany the decision to phase out nuclear energy leads to a strong demand for packages to transport low and middle active radioactive waste due to the dismantling and decommissioning of nuclear power plants. According to IAEA regulations the “non-competent authority approved package types” are the excepted packages and the industrial packages of Type IP-1, IP-2 and IP-3 and of Type A.
For the packages of Type IP-2, IP-3 and Type A an assessment by the German competent authority is required for the quality management for the design, manufacture, testing, documentation, use, maintenance and inspection. In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure.
Assessment experience on packages loaded with damaged spent nuclear fuel for transport after storage
(2018)
In 2017 the first German package approval certificate was issued for a dual purpose cask (DPC) design with encapsulated damaged spent nuclear fuel. At the Bundesanstalt für Materialforschung und -prüfung (BAM) a comprehensive assessment procedure was carried out with respect to the mechanical and thermal design, the containment design and quality assurance for manufacturing and operation. Main objective of this procedure was to verify the Package Design Safety Report (PDSR) fulfils the requirements according to the IAEA regulations SSR-6.
Until now only standard spent nuclear fuel assemblies were designated for interim storage and transports. Due to nuclear phase out in Germany all other kinds of SNF in particular damaged fuel has to be packed. Therefore specific requirements have to be considered in accordance with international experiences written in IAEA technical reports. In Germany damaged spent nuclear fuel (DSNF) needs a tight encapsulation with special encapsulations and clearly defined properties.
Due to the limited amount of DSNF these encapsulations are designed for storage and transport in existing packages. From the assessment experience it has been seen, corresponding PDSR need an extensively expansion to cover the design of these encapsulations and their influences on the package. Then such well-defined encapsulations can be handled like standard fuel assemblies. The main difference to standard package components is, encapsulations with permanent closure achieve their specified condition not after manufacturing but only during operation after loading and closing.
Thus specific handling instruction and test procedures are necessary especially for welding, where BAM is able to survey the quality of this first part of operation.
For disposal of the German research reactor of the Technical University Munich FRM II a new transport and storage cask design was under approval assessment by the German authorities. The Bundesanstalt für Materialforschung und -prüfung (BAM) assessed the mechanical and thermal package safety and performed drop tests. The activity release approaches and subjects of quality assurance and surveillance for manufacturing and operation of the package were assessed by BAM as well.
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 a wood-filled impact limiter is installed to reduce impact loads to the cask under drop test conditions. In the cavity of the cask a basket for five spent fuel assemblies is arranged.
For the safety case 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 test 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-Analysis (FEA) used for the safety analysis of the package design.
The finite-element models incorporated in the package design safety report include the cask body, the lid system, the inventory and the impact limiters with the fastening system. In this context special attention was paid to the modeling of the encapsulated wood-filled impact limiters.
Additional calculations using the verified numerical models were done by the applicant and assessed by BAM to investigate e.g. the brittle fracture of the cask body made of ductile cask iron within the package design approval procedure.
This paper describes the package design assessment from the view of the competent authority BAM including the applied assessment strategy, the conducted drop tests and the additional calculations by using numerical and analytical methods.
For disposal of the German research reactor of the Technical University Munich FRM II a new transport and storage cask design was under approval assessment by the German authorities. The Bundesanstalt für Materialforschung und -prüfung (BAM) assessed the mechanical and thermal package safety and performed drop tests. The activity release approaches and subjects of quality assurance and surveillance for manufacturing and operation of the package were assessed by BAM as well.
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 a wood-filled impact limiter is installed to reduce impact loads to the cask under drop test conditions. In the cavity of the cask a basket for five spent fuel assemblies is arranged.
For the safety case 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 test 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-Analysis (FEA) used for the safety analysis of the package design.
The finite-element models incorporated in the package design safety report include the cask body, the lid system, the inventory and the impact limiters with the fastening system. In this context special attention was paid to the modeling of the encapsulated wood-filled impact limiters.
Additional calculations using the verified numerical models were done by the applicant and assessed by BAM to investigate e.g. the brittle fracture of the cask body made of ductile cask iron within the package design approval procedure.
This paper describes the package design assessment from the view of the competent authority BAM including the applied assessment strategy, the conducted drop tests and the additional calculations by using numerical and analytical methods.
Since 2005, several European countries, coordinated by the European Association of Competent Authorities (EACA), have been developing a guide on contents and structure of the documentation demonstrating the compliance with the regulations for packages for the transport of radioactive material (package design safety report, PDSR). This guide has been periodically improved, considering feedback from Designers and authorities.
Taking into account the successful application of this guide in Europe, in 2013 the International Atomic Energy Agency (IAEA) decided to establish a similar guide as an IAEA document for promotion of worldwide use. The development of this IAEA guide started from the latest version of the European PDSR guide. In 2016/2017, during a 120-day review period, comments on the draft were received from member states and international organizations. These were incorporated into the draft in a series of meetings in 2017.
In another meeting in December 2018 the draft was updated to be in line with the latest revision of the IAEA Regulations for the Safe Transport of Radioactive Material (SSR-6).
In this process the draft has been improved significantly, regarding structure as well as implementation of a graded approach depending on the package type, and clarified.
This paper points to the major considerations in developing the guide and important improvements over the last version of the European PDSR guide.