Description of the German Concept of Final Storage Cask

Abstract

The German storage concept for the direct final storage of spent fuel assemblies from LWR reactors is described. The final storage concept is designed in such a way that it encompasses the whole spectrum of fuel elements to be stored from German reactors, Le. U fuel assemblies and MOX fuel assemblies with a mean bumup of 55 GW.d.t^?1 heavy metal were considered. The further design requirements are defined in such a way that the cask concept satisfies the conditions for type B(U) transport, interim storage and fmal storage. The safe long-term containment of the activity is guaranteed by an inner cask welded leak-tight; the sufficient shielding and the transport packaging are ensured by a shielding cask.     

Regulatory control of the German radioactive waste management system

In Germany, all radioactive materials are controlled from their origin to their final disposal or release. This is in accordance with the IAEA draft Convention on the Safety of Radioactive Waste Management. A Waste Control System has been developed and is now available. Although the IAEA and EU regulations allow disposal outside the country of origin this philosophy is not in compliance with the present German disposal concept. As a consequence, Germany will not grant licences for the import of radioactive waste for final disposal or for export for disposal in a foreign country. Thus, unconditional clearance is a prerequisite if the final destination is outside Germany. However, Germany will closely follow all international developments in the field of radioactive waste disposal.     

The German Konrad repository project

In the Federal Republic of Germany, it is planned to dispose of radioactive waste with negligible heat generation in the abandoned Konrad iron ore mine. Under the law, the Physikalisch-Technische Bundesanstalt (PTB) is the authorized applicant and thus responsible for the construction and operation of federal installations for the long-term storage and disposal of radioactive waste. On the basis of site-specific safety assessments covering the overall geological and hydrogeological situation, the technical concept of this facility including its scheduled mode of operation and the waste packages to be disposed of, the PTB has demonstrated the safety of the planned Konrad repository in the operational and post-operational phase.     

German NDE within PISC-II program

Within PISC-I, which was finished at the end of the seventies, three uncladded 200 mm thick plates were available. In this program ten European countries were participating. The test specimens came out of the American HSST-program. A series of testing according to the ASME-procedure were performed and in addition several alternative techniques. The manufacturing defects dominated but were unrealistic large. There was a lack of small cracklike defects.The PISC-II program was initiated in the early eighties with participation of twelve European countries and in addition USA, Japan and Canada. A lot of realistic service induced cracklike defects were available especially what concerned their position and size.     

Failure pressure and failure mode of the latest type of German pwr containments

For the pressurized water reactor Philippsburg (KKP II) a slowly increasing containment pressure typical for a core melt accident is assumed. The failure pressure and failure mode of the containment are determined. It turns out that the sealing box covering the bolted connection at the equipment hatch will fail at a containment over-pressure between 12.9 and 13.7 bar. Then the leakage through the bolted connection is sufficient to prevent further pressure increase. However, if the sealing box failed at a somewhat higher pressure, a global containment failure with extreme mechanical damage would have to be expected.     

COCOSYS: Status of development and validation of the German containment code system

For the simulation of severe accident propagation in containments of nuclear power plants it is necessary to assess the efficiency of severe accident measures under conditions as realistic as possible. Therefore the German containment code system COCOSYS is under development and validation at GRS. The main objective is to provide a code system on the basis of mostly mechanistic models for the comprehensive simulation of all relevant processes and plant states during severe accidents in the containment of light water reactors covering the design basis accidents, too. COCOSYS is being used for the identification of possible deficits in plant safety, qualification of the safety reserves of the entire system, assessment of damage-limiting or mitigating accident management measures, support of integral codes in PSA level 2 studies and safety evaluation of new plants.COCOSYS is composed of three main modules, which are separate executable files. These modules are covering thermal hydraulics including hydrogen combustion, fission products mainly aerosols and iodine behaviour, and corium behaviour with molten corium concrete interaction. The communication between these modules is realized via PVM (parallel virtual machine).COCOSYS is subject to an ongoing internal and external validation process. At present this validation process is mainly based on tests being performed in the German ThAI facility. Experiments to be performed in ThAI dealing with hydrogen combustion, recombiner behaviour and aerosol and iodine issues are currently subject of the just started OECD-THAI project. Examples given for the successful validation are the participation in the OECD/NEA ISP-47 and the benchmark for the CCI-2 test in the frame of the OECD-MCCI project.For example COCOSYS has been used in licensing procedure performed for the installation of catalytic recombiners in German nuclear power plants. At present COCOSYS is in use for the licensing process of the new Finnish EPR plant on the industrial side.Improvements and model extensions like pyrolysis processes, direct containment heating and the combined use with CFD models are just ongoing.     

FRM-II: The new German research reactor

A new German high-flux research reactor is presently being built in Garching by the Technical University of Munich. The new reactor, called FRM-II, shall replace the existing ‘Forschungsreaktor München' FRM which has been operating very successfully for about 40 years now. The new reactor has been optimized primarily with respect to beam tube applications of slow neutrons, but will also allow to irradiate samples with thermal neutrons. Therefore, the FRM-II has been designed to provide a high flux of thermal neutrons in a large volume outside of the reactor core, where the neutron spectrum can be locally modified by using special spectrum shifters. The goal was to further obtain this high flux at a reactor power being as low as possible since this represents the best choice because of the lowest background radiation for the experiments, the lowest nuclear risk potential, the lowest costs and superior inherent safety features. In April, 1996, the project obtained the first partial nuclear licence which covered the general acceptance of the safety concept, the site opening and the construction of the reactor building. The final partial nuclear licence which allows nuclear start-up is expected for the year 2001.     

Evaluation of piping damage in German nuclear power plants

Information relating to piping damage in safety-relevant systems in German nuclear power plants with light water reactors (both pressurized water reactors (PWRs) and boiling water reactors (BWRs)) were analyzed with respect to the modes and the causes of damages. In general, the total range of observed piping damage is low. The incidents (82) in plants with PWRs affected mainly pipes with small diameters. Almost all damaged piping showed wall-penetrating cracks combined with leakages, which revealed the damage. Initial cracks at piping with larger diameters were discovered in isolated cases during in-service inspections. With regard to the incidents (71) in plants with BWRs, piping with small as well as large diameters was affected to different degrees. Wall-penetrating cracks combined with leakages were detected at piping with small diameters. For large-diameters pipes, cracks were indicated during in-service inspections and supplementary examinations. The results of the incident evaluations confirm the conservativeness of the safety concept chosen for the design of German nuclear power plants with light water reactors.     

Operating experience with pressurised components of German light-water reactors

Operating experience of pressurised components is reported on the basis of 19 light-water reactors operating in Germany. The design basis and materials have demonstrated their worth. Licenses are not limited in time, and the major regulatory effort is directed to continuous improvement in plant safety. Technical issues for long-term operation as evaluation of operating experience, plant monitoring, replacement of components are addressed. The basic safety concept as the design basis for the pressurised components is illustrated by some of its details. The main results from the analysis of operating experience are mentioned. Plant monitoring and inspections are important measures to ensure the integrity of the components and to maintain the safety level of the plant. The expansion of the monitoring system may allow a reduction in the scope of inspections.     

Accident with Packages for RAM on a German Autobahn

Abstract

This brief note describes a car accident in which three packages for radioactive materials were involved.