Gesellschaft für Nuklear Service (Germany)

The Fairway Basin lies between Australia and New Caledonia in the northern Tasman Frontier area with water depths ranging from less than 1,000–2,400 m. This basin formed in mid-to-Late Cretaceous during eastern Gondwana breakup and received detrital and pelagic sediments since that time. It is known for a 70,000 km2 bottom simulating reflector interpreted as either one of the world’s largest gas hydrate layers or as a regional diagenetic front. Seismic reflection data shows sedimentary thicknesses (up to 4 km) and geometries capable of trapping hydrocarbons. We interpret seismic stratigraphy and available well data in terms of paleogeography and tectonic evolution. This work allowed the discovery of a deeply buried delta probably of the same type as the deepwater Taranaki Delta. This stratigraphic framework is used to constrain multi-1D generation modelling and to test three main hypotheses of source rocks. The most likely scenario, similar to the one accepted for the Taranaki petroleum province, are a type-III and type-II source rocks intercalated in Cretaceous prograding series. Another possible scenario is a source rock equivalent to the East Australian Walloon Formation and occurrence of marine source rock in the pre-rift sequence. Although large modelled volumes at this stage are speculative due to limited data on source rock composition, richness and distribution, as well as on the presence and quality of reservoir and seal, this study confirms the prospectivity of the Fairway Basin and the need for more data to further assess this basin.

In order to offer a more cost effective, safer and efficient Intermediate Level Waste (ILW) management service, EnergySolutions EU Ltd. and Gesellschaft fur Nuklear-Service mbH (GNS) have been engaged in the development of integrated radioactive waste retrieval, packaging and conditioning solutions in the UK. Recognising the challenges surrounding regulatory endorsement and on-site implementation in particular, this has resulted in an alternative approach to meeting customer, safety regulator and disposability requirements.By working closely with waste producers and the organisation(s) responsible for endorsing radioactive waste management operations in the UK, our proposed solutions are now being implemented. By combining GNS’ off-the-shelf, proven Ductile Cast Iron Containers (DCICs) and water removal technologies, with EnergySolutions EU Ltd.’s experience and expertise in waste retrieval, safety case development and disposability submissions, a fully integrated service offering has been developed. This has involved significant effort to overcome technical challenges such as on-site equipment deployment, active commissioning, conditioning success criteria and disposability acceptance.Our experience in developing such integrated solutions has highlighted the importance of working in collaboration with all parties to achieve a successful and viable outcome. Ultimately, the goal is to ensure reliable, safe and effective delivery of waste management solutions.Copyright © 2013 by ASME

During the operation and dismantling of nuclear plants, a variety of radioactive liquid wastes accumulate, which must be supplied to a disposal chain suitable for final disposal. In the initial conditioning step, with regard to their chemical properties these waste waters are usually adjusted by water treatment systems (predominantly evaporation facilities) permanently installed in nuclear power plants, and then, they are concentrated up to a predefined content of solids. Afterwards, the developing evaporator concentrate is desiccated locally and suitable for final disposal by own permanently installed conditioning installations or by means of mobile plants. For a mobile conditioning of evaporator concentrates, GNS uses the drying plant of the type FAVORIT proven over decades. Each of the mobile FAVORIT processing plants is provided with a handling licence for Germany valid nationwide according to § 7 of the German Radiation Protection Ordinance (StrlSchV).

This paper describes the procedure for the qualification of large components (Steam Generators) as an IP-2 package, the ship transport abroad to Sweden and the external treatment of this components to disburden the Nuclear Power Plant from this task, to assure an accelerated the deconstruction phase and to minimize the amount of waste.

The evaluation of fracture mechanics properties at intermediate strain rates can be carried out on the basis of different standards. The starting point of the lecture is a comparative evaluation of the relevant standards and test procedures [1, 2, 3, 4]. The aim of the fracture mechanics test at intermediate strain rates are not only generally materials characterizations but often the portability of the determined parameters on bigger sizes in case of evaluation of real components. In this situation the definition of component transferable material parameters is an important problem. In this context the crack initiation point gets an all-dominant importance [5, 6]. In case of multiple specimens testing typically the definition of the crack initiation point based on the results of stretch-zone measurements.

Final disposal casks will be stored up to 60 years in an intermediate storage facility. Unauthorized access to the radioactive content has to be detected. For this reason a short weld seam acting as a seal is applied between the cask body and the lid. The seal has to be examined using unique and reproducible features characterizing the weld seams even under the influence of aging. Relief inspection is investigated for this task. Received images are segmented into slices corresponding to different altitude levels. Every slice shows a number of objects, which are analyzed by morphometrical and contour features. The identification algorithm is basically a supervised minimum distance classifier. The initial references are learned automatically by the system when a cask is introduced into the storage. References will be updated every time a cask is inspected.

During the interim storage of spent fuel rods the verification of the identity as well as of the integrity of special shielding casks containing final disposal packages has to be guaranteed by applying appropriate safeguards measures over a period of up to 50 years until a geological repository goes into operation. The casks to be verified are made of cast iron (GGG 40) and consist of the actual final disposal package and a cylindrical shielding cask with a height of about 6 m and a diameter of about 1.5 m which is closed with a screwed-on lid. It is planned to link the lid to the cask by a short weld seam, so that the lid can only be removed by destroying the weld. The identification of the weld allows for statements on the integrity of the cask as well as on its identity. The latter is, in addition, supported by non-reproducible cask characteristics.