Does gentle regulation have a chance?

The Repository Site Selection Act can only be successfully implemented when it is embedded in a comprehensive context of gentle regulation. The Repository Site Selection Act accentuates not only options for solving conflicts prior to judicial legal protection. The Act also furthermore features supplementary, informal forms of public participation. Diverse examples of encounter, contention and discourse, all of which are resources for solving conflict, as forms of gentle regulation are presented (national support board, participation representative, development of public participation, particularly informal public participation, compensation and socioeconomic development, political communication and storytelling, restorative justice, historical analyses, narrative salon). This perspective has so far had very little discussion. The (legal) discussion is concentrated on possibilities for legal protection in court. The attempt at gentle regulation needs further mediation and discussions if the innovative approach in the repository site selection procedure should be successfull. The starting point is a jurisprudence, which questions success and effectiveness of regulations and discusses various ways and approaches.

What happens when the nuclear power plant goes? Social and economic consequences of dismantling for the site and its population

This project deals with the question of what the overall social and economic consequences of dismantling a nuclear power station are for the population and the site. Various disciplines and specialist fields are concerned with questions that touch on the topic of dismantling nuclear technical facilities; however, there are so far no research projects that examined these processes from social scientific, geographic and engineering scientific perspectives. This article concentrates predominantly on the former perspective of the dismantling. Within the framework of this project the affected population and experts from the communities were asked how they deal with the dismantling of the nuclear power stations, which were triggered by the rapid change in energy policy following the accident in Fukushima in 2011. It became clear that there were various strategies for dealing with this process depending on the location. This was the reason to follow up the question of coping with this process at different locations. It could be shown, for example, that the consequences of this event were essentially determined by how the community was already positioned beforehand, e.g. whether the economic situation was a monostructure or whether long-term considerations about the future had already been made during the operating time of the power station. At the individual level, the “prerequisites” in the sense of individual value orientation and the spatially related identity, were also essentially responsible for how the risks of the dismantling and the further development of the community were perceived and evaluated. Furthermore, it was compiled from where the people extracted their information, which sources had a high or low credibility, which worries they have with respect to the near future and whether they have the intention to leave the community. In this project it became clear that there were examples of best practice with respect to dealing with this rapid and fundamental change at the locations.

State-of-Knowledge and Guidance in EURAD Knowledge Management (Work Packages 11 State-of-Knowledge & 12 Guidance)

EURAD, the European Joint Programme on Radioactive Waste Management (RWM), is the European Research Programme on RWM, aimed at supporting member states with the implementation of their national programmes. It brings together over 100 organisations from different backgrounds and countries, which work together in RD&D projects, Strategic Studies and Knowledge Management (KM). The importance of KM is recognised by EURAD and reflected in a number of activities. One essential activity is the capture of the current State-of-Knowledge in the field of RWM and its transfer to the implementation of the different national programmes. This is done by different types of Knowledge Documents that are made available through a dedicated IT tool (e.g. a Wiki). The development of the individual EURAD KM documents is performed by recognised experts. These experts will share their view on the most relevant knowledge on a specific topic, highlighting safety functions and operational aspects. Additionally, signposting to pre-existing documents is performed (State-of-the-Art Documents, Scientific Papers, etc.). The hierarchy of the works for the KM documents (Theme Overview, Domain Insight, State-of-Knowledge, Guidance) is closely linked to the generic EURAD Roadmap/GBS (Goals Breakdown Structure). It provides a hierarchical structure that facilitates definition, organisation and communication of topics. All of this allows knowledge to be captured and presented with the level of detail that is required by the end-user, from a broad overview down to an increasing level of detail (pyramid of knowledge). To ensure the quality and consistency of the documents with the overall EURAD KM approach, quality assurance and editorial procedures are applied. Collection of end-user feedback will aid the optimisation and further development of the KM activities. To facilitate the transfer of knowledge, the EURAD KM programme goes beyond documents and strives to facilitate exchange between people and signpost to other resources, such as Training and Mobility activities (also organised by EURAD Work Package 13 Training & Mobility) or Communities of Practice. All these activities will contribute to a useful and end-user-friendly EURAD KM programme that is designed to be operational well beyond the runtime of EURAD-1. This presentation will provide further insight into the approaches, status of work and an outlook on future activities that will support member states with the implementation of their national programmes.

Spectroscopic and chemical investigations on volatile fission and activation products within the fuel-cladding interface of irradiated pressurised water reactor fuel rod segments

In Germany, the present waste management concept foresees the direct disposal of spent nuclear fuel (SNF) in deep geological repositories for high-level waste available by 2050, at best. Until then, SNF is encapsulated in dual-purpose casks and stored in dry interim storage facilities. Licenses for both casks and facilities will expire after 40 years following loading of the cask and emplacement of the first cask in the storage location. Yet, due to considerable delays in the site selection process and the estimated duration for construction and commissioning of a final repository of at least 2 decades, a prolonged dry interim storage of SNF is inevitable (ESK, 2015). Concerning these considerable timespans, integrity of the cladding is of utmost importance regarding the ultimately conditioning of the fuel assemblies for final disposal. Various processes strain the structural integrity of Zircaloy cladding during reactor operation and beyond such as delayed hydride cracking, fuel-cladding chemical interactions or irradiation damage induced by α-emitters present in the fuel pellet's rim zone (Ewing, 2015). Especially with higher burn-up, the gap between fuel and cladding closes and results in the formation of an interaction layer, in which precipitates of fission and activation products are present, displaying an interface for degradation processes. For chemical analysis and speciation of these agglomerates, Zircaloy-4 and SNF specimens were sampled from fuel rod segments irradiated in commercial pressurised water reactors during the 1980s. Zircaloy-4 specimens were taken from an UOX (50.4 GWdtHM-1) and mixed oxide fuel (MOX) (38.0 GWdtHM-1). In addition, SNF fragments were sampled from the closed gap of both fuel types to examine volatile activation and fission products, which had been segregated from the centre to the pellet periphery during irradiation and thus contribute to the possible chemically assisted cladding degradation effect of the precipitates within the fuel-cladding interface. Spectroscopic analysis of precipitates within the interface layer between fuel and cladding were performed by optical microscopy, X-ray absorption and X-ray photoelectron spectroscopy, as well as by energy-dispersive scanning electron microscopy. Moreover, the radionuclide inventory of the respective Zircaloy-4, fuel and interaction layers was determined using liquid scintillation counting, γ-spectroscopy, gas mass spectrometry, ion chromatography and inductive-coupled plasma mass spectrometry and compared to results received by MCNP/CINDER and webKORIGEN calculations. In this study, we provide results regarding the speciation and chemical composition of previously identified Cs-U-O-Zr-Cl-I bearing compounds found in the interaction layer of irradiated nuclear fuel and inventory analyses of radionuclides present therein, with particular emphasis on Cl-36 and I-129. Furthermore, the agglomerates within the fuel-cladding interface were characterised for the first time utilising synchrotron radiation-based Cl K-edge and I K-edge measurements, resulting in compounds with structural similarities to CsCl and CsI. The outcomes obtained from this study provide further insights into the complex chemistry within the fuel-cladding interface with respect to the aging management and integrity of SNF under the conditions of interim storage. In future studies we will examine whether the different compounds at the fuel-cladding interface have the potential to affect the mechanical properties of Zircaloy cladding.

PREDIS: innovative ways for predisposal treatment and monitoring of low and medium radioactive waste

The EURATOM PREDIS project (http://www.predis-h2020.eu, last access: 25 October 2021) targets the development and implementation of activities for predisposal treatment of radioactive waste streams other than nuclear fuel and high-level radioactive waste. It started on 1 September 2020 with a 4 year duration. The consortium includes 47 partners from 17 member states. The overall budget of the project is EUR 23.7 million, with EC contribution of EUR 14 million. The PREDIS project develops and increases the technological readiness level (TRL) of treatment and conditioning methodologies for wastes for which no adequate or industrially mature solutions are currently available, including metallic materials, liquid organic waste and solid organic waste. The PREDIS project also develops innovations in cemented waste handling and predisposal storage by testing and evaluating. The technical work packages align with priorities formulated within the Roadmap Theme 2 of EURAD (https://www.ejp-eurad.eu/sites/default/files/2021-09/2_Predisposal_Theme_Overview.pdf, last access: 15 October 2021), Nugenia Global Vision (https://snetp.eu/wp-content/uploads/2020/10/Global-vision-document-ves-1-april-2015-aa.pdf, last access: 15 October 2021) and with those identified by the project's industrial end users group (EUG). The PREDIS will produce tools guiding decision making on the added value of the developed technologies and their impact on the design, safety and economics of waste management and disposal. Four technical work packages are focusing on specific waste types: metallic, liquid organic, solid organic, and cemented wastes. For the first three, the main aim lies in processing, stabilizing, and packaging the different waste streams, e.g. by using novel geopolymers, to deliver items which are in line with national and international waste acceptance criteria. In contrast, the fourth technical work package has a different focus. To provide better ways for a safe and effective monitoring of cemented waste packages including prediction tools to assess the future integrity development during predisposal activities, several digital tools are evaluated and improved. Safety enhancement (e.g. less exposure of testing personnel) and cost-effectiveness are part of the intended impact. The work includes but is not limited to inspection methods, such as muon imaging, wireless sensors integrated into waste packages as well as external package and facility monitoring, such as remote fiber optic sensors. The sensors applied will go beyond radiation monitoring and include proxy parameters important for long-term integrity assessment (e.g. internal pressure). Sensors will also be made cost-effective to allow the installation of many more sensors compared to current practice. The measured data will be used in digital twins of the waste packages for specific simulations (geochemical, integrity) providing a prediction of future behavior. Machine learning techniques trained by the characterization of older waste packages will help to connect the models to the current data. All data (measured and simulated) will be collected in a joint database and connected to a decision framework to be used at actual facilities. The presentation includes detailed information about the various tools under consideration in the monitoring of cemented waste packages, their connection and first results of the research.

Development of a database for the analysis of the disposal system in the representative preliminary safety analysis

The Federal Company for Radioactive Waste Disposal (BGE) is the German waste management organisation responsible for implementing the search for a site with the best possible safety for the disposal of high-level radioactive waste for at least 1 million years, following the amendments of the Repository Site Selection Act in 2017. The selection procedure is meant to be a participatory, transparent, learning and self-questioning process based on scientific expertise. It consists of three phases with an increasing level of detail. The first step of the first phase of the site selection procedure was completed in September 2020 and resulted in the identification of 90 subareas that give reason to expect favourable geological conditions for the safe disposal. The potentially suitable subareas cover approximately 54 % of Germany and are located in three different host rocks: rock salt (halite), claystone and crystalline rock. The second step of phase one is currently in progress and includes the so-called representative preliminary safety analyses that aim to assess the extent to which the safe containment of the radioactive waste can be expected. Representative preliminary safety analyses are one of the foundations for deciding whether an area will be considered for surface-based exploration in the next phase of the site selection procedure. Within the preliminary safety analyses, the behaviour of the disposal system is analysed in its entirety, across all operational phases of the repository and under consideration of possible future evolution of the disposal system with respect to the safe containment of the radioactive waste. The development of a database is described, which aims to systematically document and provide the framework needed for the analyses of the disposal systems in the subareas regarding the safe containment of the radionuclides over the assessment period of 1 million years. This database includes the vast amount of information about the different components of the disposal system. This includes also the geological setting, the technical conception of the repository and compilations of values for the physical, geoscientific, and technical parameters characterising the various barriers of the disposal system. Furthermore, a self-contained derivation of expected and deviating future evolution of the disposal system and its geological setting is included; following the so-called features, events and processes (FEP) strategy.

Modelling of hydrodynamic and solute transport with consideration of the release of low-level radioactive substances

When nuclear power plants are dismantled, only a small portion is heavily contaminated with radioactivity and must be stored in a repository. The remaining material, mainly concrete rubble (construction waste), is decontaminated if necessary and can be stored in conventional surface landfills after clearance. The focus of this work is on the modelling of such landfills and the radioactive substances during raining events. The influence of the heterogeneous nature of the construction rubble should also be investigated. The simulation environment DuMux, mainly developed by our institute, is used to compare different modelling approaches. It follows a previous work by Merk (2012). The research work is supported and accompanied by the Federal Office for Radiation Protection (BfS). Parts of the research initiatives of the BfS in the area of clearance of materials with negligible radioactivity are also presented.

The anisotropic properties of granites – effects of tectonic emplacement mode on potential crystalline host rocks for nuclear waste deposits in Germany

For permanent nuclear waste disposal sites, crystalline rocks, especially granitic/granodioritic batholiths, are considered an appropriate host rock. Principally, three types of granitic plutons occur in the extra-alpine crystalline basement of Germany that were consolidated during the late Paleozoic Variscan orogeny of Central Europe: (i) Pre-Variscan voluminous granodiorites that are hardly affected by the subsequent continent–continent collision; (ii) voluminous granites in various tectonic settings intruded during the late orogenic stage of the Variscides; (iii) post-orogenic granites related to vast Permian intracontinental extension. Thus, in terms of the syn-intrusive tectonic setting and post-intrusive processes there are significant differences. Although it can be expected that different tectonic environments caused significant differences in the material properties, for Germany, however, there is no systematic study regarding the fabric of such plutonites. In order to find the most suitable “granite” we investigate the primary anisotropy of granites evolved during the emplacement and crystallization of the melt. For this we sample rocks of all three principal types and various syn-intrusive tectonic settings, i.e., compression, extension, strike-slip, transtension, and transpression. By means of combined measurements of the “Anisotropy of the Magnetic Susceptibility” and the “Shape Preferred Orientation” we characterize the syn-intrusive flow pattern, i.e., the magmatic foliation and lineation. The Crystallographic Preferred Orientation is analyzed by a combination of neutron time-of-flight experiments and electron backscatter diffraction measurements at the Frank Laboratory of Neutron Physics at JINR, Dubna, Russia, and the TU Bergakademie Freiberg respectively. Furthermore, special attention is given to the systematic mapping of annealed microcracks evolved during late magmatic fluid escape and/or post-crystallization hydrothermal activity. In a second step we compare the primary anisotropy with the post-magmatic fracture pattern of the particular granites. Those fractures constitute probable fluid pathways and, thus, the first-order risk for a potential permanent nuclear waste disposal. All datasets are organized in a Geological Information System allowing for a complete traceability of the different investigation steps. The results of this study will serve as a basis for a future detailed exploration.

Site selection for the best clay-hosted repository in Switzerland

The site selection process for the deep geological repositories in Switzerland is in its final phase. All three remaining sites fulfil the requirements of the repositories for low level waste (LLW) as well as for high level waste (HLW) and spent fuel. Using surface-based exploration methods, including 3D seismic studies and deep bore holes, Nagra has recently collected the necessary data to select the most suitable site. The sites will be compared based on 13 technical criteria defined in 2008 and already used in the 2 previous phases of the selection process. The criteria comprise safety-related aspects including the barrier properties and their long-term stability as well as the construction suitability of the repository and its access facilities. If one site offers optimal conditions for both HLW and LLW and the space requirements can be met at that site, a co-disposal facility will be proposed. This facility will then comprise separate emplacement areas with specific safety concepts for the different waste categories. In our contribution we present the overall approach for the surface-based exploration in the different stages of the site selection process. We show how we integrated third party data, seismic surveys, surface mapping as well as deep and shallow bore holes to drive the site selection process. Example data sets from the ongoing deep drilling campaign (clay content, hydrotest data, porewater isotope profiles etc.) and recent 3D seismic surveys are used to illustrate our approach to distinguish the remaining sites according to long-term safety and to underline the large safety margin of the selected clay host rock in long-term evolution scenarios.

Development of a mobile, automated, optical inspection system for radioactive barrels

Due to the delayed construction and commissioning of a German repository for intermediate- and low-level radioactive waste, waste inventories from several decades are now located at the interim storage sites, the safe custody of which must also be ensured for an indefinite period of interim storage. The usual practice in the interim storage facilities is recurrent inspections, which are carried out almost exclusively manually and without electronic comparative recordings as well as without mechanical documentation and archiving. Remote or automated inspection does not take place. The inspections are carried out visually and are therefore very subjective and thus subject to errors. Manual performance is labor intensive and requires the use of personnel exposed to radiation. Neither are uniform inspection criteria of the visual inspections applied, nor are the inspections performed uniformly between sites. Based on these facts, the Institute for Technology and Management in Construction, Department of Deconstruction and Decommissioning of Conventional and Nuclear Buildings, together with the Institute for Photogrammetry and Remote Sensing, is developing an automated drum inspection system as part of the funding measure FORKA – Research for the Deconstruction of Nuclear Facilities. EMOS is a mobile inspection unit that remotely and automatically records the entire surface of the drum, including lid and bottom, optically; evaluates it analytically; and both stores it electronically and outputs the results in the form of an inspection report. In this way, recurring inspections of the drum stock can be completed under the same inspection conditions each time. A decisive advantage is the possibility of carrying out the inspection remotely in order to reduce the radiation dose to the employees on site. The optical evaluation, display and output of the results will ensure a more precise inspection and analysis of the drum surfaces through software to be specially developed than is possible through manual and visual inspections as currently performed in the interim storage facilities. The continuous monitoring of the stored drums will be facilitated and also the tracing of possible damage development through the comparison of archived measurement results is a novel and powerful tool that helps to increase and ensure the safety aspects of interim storage in the long term. Changes in drum geometry as well as in the surface condition (e.g. corrosion formation, etc.) can be identified at an early stage with the help of the inspection unit, and measures can be taken at an early stage to counteract the loss of integrity of the storage containers.