Apresentamos um estudo de solidificaçaõ e lixiviacão de matrizes de vidros porosos de alto teor de sílica armazenando 20% em peso de solução aquosa simuladora de rejeitos nucleares de alto nível de radioatividade do tipo Savanah River Labratory. A matriz porosa foi preparada após o tratamento térmico de um vidro de 65% SiO2-27%B2O3-8%Na2O, que sofreu separação de fase à 560°C por 20 horas e lixiviado em HCl - 3N à 90°C. O tamanho dos poros (tipicamente de 100 à 250Å de diâmetro) , foi determinado utilizando o método BET. Após sinterização à 1300°C em ar, as amostras foram caracterizadas física e quimicamente através de testes de lixiviação padronizados MCC1, Soxhlet (MCC5) e Estagnante durante cerca de 28 dias. Determinamos a perda de peso total, o pH, as taxas de lixiviação diferencial e as concentrações acumuladas para os seguintes elementos: Si, Na, B, Ca, Mn, Al, Fe e Ni com técnicas de ICP e espectroscopia de chama para o caso do Na. Os resultados são comparados com os obtidos com vidros borosilicatos de referência, fabricados por fusão convencional (SRL 131, PNL 76-68, MCC 76-68, SRL TDS 131, AVM-Ml à M7), vidros fabricados pela técnica sol-gel (TDS 211), vidros de alto teor de sílica (CU PGM), synroc-D, cerâmicas manufaturadas, concreto FUETAP e matrizes metálicas. Os valores obtidos são similares àqueles obtidos com os melhores vidros borosilicato presentemente usados. / We present a study of the sinterization and of the leaching behavior of a high silica porous glass matrix containing 20 weight % of simulated solution of high level liquid nuclear waste of the type Savanah River Laboratory. The porous matrix has been prepared after heat treatment of a 65% SiO2-27%B2O3-8%Na2O glass, phase separate at 560°C for 20 hours and leached in 3N HCl at 90°C. The pore size (typically 100-250Å in diameter) has been determined by the BET method. After sinterization in air at 1300°C, the samples have been physically and chemically characterized during 28 days using the MCC1, Estagnant and Soxhlet (MCC5) standard tests. We have determined thetotal weight loss, the pH, the diferential leaching rate and the cumulative concentrations for the following elements: Si, Na, B, Ca, Mn, Al, Fe and Ni by ICP technique, for Na flames spectroscopy. The results are compared with these obtained with other reference borosilicate glasses made by conventional fusion techniques (SRL 131, PNL 76-68, MCC 76-68, SRL TDS 131, AVM-M1 to M7), glasses made by sol-gel technique (TDS-211), porous glasses matrix (CU PGM), synroc-D, tailored ceramics, FUETAP concrete and metallic matrix. The values obtained are similar to those found for the best borosilicate glass presently used.

Made available in DSpace on 2014-10-09T12:35:13Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T13:57:01Z (GMT). No. of bitstreams: 0 / As fontes radioativas seladas em desuso, descartadas como rejeito radioativo, constituem uma parcela dos rejeitos radioativos que merece atenção especial, por sua atividade possuir potencial para causar doses de radiação elevadas, em indivíduos inadvertidamente expostos. Já é significativo o volume desses rejeitos. Manter essas fontes armazenadas em depósitos provisórios, indefinidamente, seria transferir o problema às futuras gerações. O presente estudo propõe o uso de indicadores de segurança complementares à dose e risco para o desenvolvimento de uma metodologia de avaliação da segurança de depósitos finais destinados à deposição de fontes radioativas seladas que demonstre que o isolamento será suficientemente seguro pelo tempo necessário para obter a licença da instalação, com custo acessível aos países em desenvolvimento. / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Os nanotubos naturais da paligorsquita, por apresentarem propriedades físicas e químicas específicas, têm potencial uso como nano sorventes e matrizes para imobilização, retenção, e solidificação de radionuclídeos presentes em efluentes nucleares. No processo de desenvolvimento de materiais com propriedades de sorção visando a incorporação e imobilização de radionuclídeos, as etapas mais importantes são a geração de sítios ativos simultaneamente com o aumento da área superficial específica e tratamento térmico adequado para conduzir ao colapso estrutural. Neste estudo foram avaliados parâmetros e condições determinantes no processo de ativação dos nanotubos naturais da paligorsquita visando a sorção de radionuclídeos de interesse na estrutura dos nanotubos e a avaliação posterior dos parâmetros que afeitam ao colapso estrutural por tratamento térmico. Por este estudo constatou-se que a otimização do processo de ativação ácida é fundamental para o aumento da capacidade de sorção de níquel usando estruturas de nanotubos naturais ativados. A condição otimizada de ativação superficial, mantendo a integridade estrutural foi removido cerca de 33,3% dos cátions de magnésio, equivalente a 6,30·10-4 mol·g-1 de magnésio em massa, aumentando a área superficial específica em 42,8%. Este aspecto permitiu a incorporação de mesma concentração molar de níquel presente nos rejeitos radioativos líquidos em um tempo de processo de 80min. / Natural palygorskite nanotubes show suitable physical and chemical properties and characteristics to be use as potential nanosorbent and immobilization matrix for the concentration and solidification of radionuclides present in nuclear wastes. In the development process of materials with sorption properties for the incorporation and subsequent immobilization of radionuclides, the most important steps are related with the generation of active sites simultaneously to the increase of the specific surface area and suitable heat treatment to producing the structural folding. This study evaluated the determining parameters and conditions for the activation process of the natural palygorskite nanotubes aiming at the sorption of radionuclides in the nanotubes structure and subsequent evaluation of the parameters involve in the structural folding by heat treatments. The optimized results about the maximum sorption capacity of nickel in activated natural nanotubes show that these structures are apt and suitable for incorporation of radionuclides similar to nickel. By this study is verified that the optimization of the acid activation process is fundamental to improve the sorptions capacities for specifics radionuclides by activated natural nanotubes. Acid activation condition optimized maintaining structural integrity was able to remove around 33.3 wt.% of magnesium cations, equivalent to 6.30·10-4 g·mol-1, increasing in 42.8% the specific surface area and incorporating the same molar concentration of nickel present in the liquid radioactive waste at 80 min.

Dans le concept actuel du stockage géologique des déchets radioactifs en France, les interfaces entre la roche encaissante, une argilite, et les matériaux cimentaires utilisés pour les bouchons de scellement et les corps des alvéoles de stockage pourraient subir une température de 70°C due à l’activité exothermique de déchets. Les évolutions minéralogiques, microstructurales et leurs conséquences sur les propriétés de transport à ces interfaces sont mal connues dans ces conditions de température.Deux dispositifs expérimentaux sont conçus. Le premier consiste à créer des interfaces pâte de ciment CEM I / argilite de Tournemire en cellules de diffusion. La chimie des solutions est suivie dans le temps et quatre échéances permettent d’étudier l’évolution temporelle des matériaux. Le second dispositif consiste à créer de telles interfaces in situ à 70°C dans le laboratoire souterrain de Tournemire. Ce dispositif, plus représentatif des conditions de stockage, est démantelé après un an d’interaction. Au préalable, le comportement de la pâte de ciment CEM I à l’issue d’une augmentation de température de 20 à 70°C est analysé. La modélisation en transport réactif (Hytec) est utilisée en support à la compréhension des évolutions physico-chimiques.La néoformation de tobermorite, de phillipsite (in situ uniquement), de C-A-S-H et de calcite formant un ruban à l’interface est avérée. Une cinétique de précipitation de la tobermorite a ainsi pu être évaluée. La pâte de ciment est décalcifiée et carbonatée. La porosité totale diminue dans la pâte de ciment, malgré une ouverture de la macroporosité par dissolution de portlandite. L’argilite semble être peu altérée. La température accélère la diffusion, tandis que les variations de porosité et le ruban ne changent pas significativement les propriétés de diffusion sur une année. / Radioactive wastes in future deep geological disposals will generate heat and locally increase temperature in the engineered barriers and host-rock. In the French design of disposal cells, temperature may reach 70°C in cementitious materials and at their contact with the clayey host-rock. The impact of temperature under such disposal conditions is still poorly known, especially regarding the geochemical and physical evolution at the interface between these two materials.Two experimental devices are designed. The first involves creating interfaces between OPC paste and argillite of Tournemire in diffusion cells. The evolution of solutions and materials are analysed over time. The second device involves creating OPC paste / argillite interfaces at 70°C under in situ conditions in the underground laboratory of Tournemire (France). This device, more representative of a deep disposal, is dismantled after one year. Prior to interface study, behaviour of the OPC paste after a temperature increase from 20 and 70°C was analysed and simulated. Reactive transport modelling supports the experimental results in order to better understand the physico-chemical evolutions at the interface.Neoformation of tobermorite (well-crystallised C-S-H), phillipsite (only in situ), C-A-S-H and calcite formed a layer at the interface. A kinetic of tobermorite precipitation is evaluated. Significant decalcification and carbonation were noticed in the cement paste. Total porosity decreases in the cement paste despite an opening of the macroporosity due to portlandite dissolution. Argillite seems to be weakly altered even if alkaline plume goes deeply through it. Porosity changes do not alter significantly diffusive properties at the studied time scale.

On the current state of material distribution of renewable energy plans for development combined with the alternative uses of innovative technologies, there have been multilateral institutional partnerships regulating the actual distribution of nuclear energy programs through the PPPs, which have maintained a primarily scientific role while attracting international attention. In addition, the specific combination of scientific knowledge and technology transfers associated with public-private regulatory spheres has led to a common co-evolution of essential development characteristics, which have been intertwined with public environmental programs and resulting activities referring to the nuclear risk management of nuclear power plants NPPs, and to the formulation of participatory protection mechanisms. In this study, I analyze the comparative institutional status of nuclear energy models in industrial transition stages with waste disposal systems which have been based in Canada and Taiwan. The research focus in this dissertation has been placed over the practical need to identify the adaptive policy approaches in governance leading to local territorial interactions interrelated with a contemporary escalation of environmental technology issues, associated with public-private partnerships (PPPs), especially in terms of operability of STS transfers (science, technology, and societies) developed at societal level. Structurally speaking, the first section of this dissertation discusses introductory explanations already presented in July 2016 for the university commission about the proposed doctoral research design. The second and final parts of this dissertation have been developed at length in view of exploring some of the issues concerning the STS energy transfers and NPPs research policies associated with PPPs configurations. The final discussion section will summarize the literature findings about the changing mechanisms established in energy governance. The evaluative findings have been mostly developed through library archival documents, national reports, and analytical studies which I have compared in this dissertation. Overall for starting point, it can be affirmed that a technocratic vision of dynamic disciplinary elements related to managerial energy configurations of nuclear power plants, including waste disposal programs, has been proposed at regional level through common identification systems, established over public provisions involving regulatory interactions of nuclear sector industries based in East Asia and Canada. International and national attention has been focused on environmental cases of post-disaster emergencies and risk protection factors, particularly following on the Fukushima nuclear plant crisis in Japan in 2011. This structural process has been classified as an international critical domain. Essentially, the constructive experience acquired in governance has relied on cross-countries interpretative democratic models based on the existence of collective information i exchanges, which have actually involved different national regulators, public development actors, and industrial management partners, supported by: scientific experts, regional state officials, non-governmental representatives, and local district communities, among others. Moreover, the resulting collaboration process for public regulatory implementation which has been followed according to governmental aims and rationalization of resources regarding the civilian nuclear energy activities has also acquired a divergent character identified in multi-level state distribution systems. This happens in view of the similar formulation of industrial transition incentives for innovation and technology transfers, also entailing attentive responses formulated by taking into account the material normative reflections; which need to promote a broader view on collective participatory models, also based on public consensus criteria. Consequently, it can be considered that nuclear energy technologies and industrial knowledge transfers have been interlinked to a public set of normative appeals and confidence measures, promoting fundamental support for governance integrative practices. From an industrial point of view, the differentiation of innovation systems pursued through the development of specialized technology districts, for instance, in East Asia and Europe, has been configured according to public-private negotiation patterns assisting on the evolution of STS assessment programs. The corresponding formulation of risk prevention measures and safety assessment principles has been addressed according to the transition obtained with the adoption of alternative renewable energy plans. Managerial innovation capacities have reflected the temporal adaptation to development changes, which have been related to the emergence of nuclear fuel-cycle radioactive programs, and nuclear waste disposal activities. At implementation level, the direct involvement of community actors and environmental institutions has come into play leading to the identification of multilevel governance routes, by enhancing the knowledge transfers and learning systems, compatible with national and local collectivities, as well as, territorial and internal capacities. At the same time, the spatial regulatory requirements for regional identifications of the technologies used and the PPP agreements prepared in connection with nuclear energy facilities, and civilian energy installations, have testified the need to introduce learning cooperation stages for the evaluative and monitoring processes. These changing adaptation stages have been publicly controversial. At the end of bitter regional local disputes, the investigative agencies producing case-based reports have indicated the status of public concern and risk perceptions on nuclear safety issues, particularly for the local population living in proximity to NPPs, reflecting on common detrimental effects in terms of public governance and mutual trust conditions. The complex variation of public understanding about the programmatic issues surrounding nuclear science development and the environmental impacts has drawn us to an analytic core of ii structural determinants, which have been investigated in order to compare the international cooperation principles and the practical nationally-based conducts. For the identification of risk protection assessments of national capacities, I have elaborated this study project for comparative purposes, by trying to emphasize the critical aspects of public STS maintenance systems, which will require a legal status and clarification for the future generations in order to guarantee security and safety for everyone.

En France, 23 000 tonnes de graphites irradiés générés par le démantèlement des réacteurs nucléaires de première génération Uranium Naturel-Graphite-Gaz (UNGG) sont en attente d'une solution de gestion à long terme. Cette thèse porte sur le comportement du tritium, l'un des principaux contributeurs à l'inventaire radiologique des graphites à l'arrêt des réacteurs. Afin d'anticiper des rejets de tritium lors du démantèlement ou de la gestion des déchets, il est indispensable d'obtenir des données sur sa migration, sa localisation et son inventaire. Notre étude repose sur la simulation du tritium par implantation de l'ordre de 3 % at. de deutérium jusqu'à environ 3 μm dans un graphite nucléaire vierge. Celui-ci a ensuite subi des recuits jusqu'à 300 h et 1300 ° C sous atmosphère inerte, gaz caloporteur UNGG et gaz humide, dans le but de reproduire des conditions proches de celles rencontrées en réacteur et lors des opérations de gestion des déchets. Les profils et la répartition spatiale du deutérium ont été analysés via la réaction nucléaire 2H(3He,p)4He. Les principaux résultats montrent un relâchement thermique du deutérium se produisant selon trois régimes contrôlés par le dépiégeage de sites superficiels ou interstitiels. L'extrapolation des données au cas du tritium tend à montrer que son relâchement thermique en réacteur pourrait avoir été inférieur à 30 % et localisé à proximité des surfaces libres du graphite. L'essentiel de l'inventaire en tritium à l'arrêt des réacteurs serait retenu en profondeur dans les graphites irradiés, dont la décontamination nécessiterait alors des températures supérieures à 1300 °C, et serait plus efficace sous gaz inerte que sous gaz humide / In France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 °C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2H(3He,p)4He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be located close to the graphite free surfaces. Consequently, most of the tritium inventory after reactor shutdown could be trapped deeply within the irradiated graphite structure. Decontamination of graphite waste should then require temperatures higher than 1300°C, and would be more efficient in dry inert gas than in humid gas

La thèse traite de la valorisation comme un processus à la fois continu, constructif et contesté. Elle aborde cette question théorique au travers de l'étude d'un objet de valorisation complexe : le traitement des déchets nucléaires aux États-Unis. Objet ontologiquement ambivalent, les déchets nucléaires peuvent être valorisés ou dévalorisés selon des perspectives très variées, qui offrent à la sociologie de l'évaluation un banc d'essai pertinent. La thèse s'appuie sur l'étude d'une série de situations où la formulation des valeurs du nucléaire pose problème. Ce sont, notamment, l'ordonnancement et les vicissitudes du processus budgétaire mis en œuvre pour financer le programme de gestion des déchets nucléaires Nord-américain, les plans conçus pour répartir la responsabilité financière dans le traitement d'un matériau qui selon toutes prévisions restera dangereux pendant un million d'années, les efforts d'un groupe d'acteurs pour attacher l'avenir de leur communauté à celui des déchets nucléaires, ou les utilisations d'une convention pour estimer la valeur économique du combustible usé. La thèse montre que les processus de valorisation et dévalorisation sont loin de se limiter à l'objet soumis à évaluation ; ces processus rendent d'autres valeurs explicites, les valeurs "de l'État", d'une communauté politique, d'une convention économique. La thèse propose d'utiliser la notion de « révaluation », d'une part, parce qu'elle permet d'expliciter l'imbrication des relations entre les processus d'évaluation et de valorisation et, d'autre part, parce qu'elle permet de souligner la particularité de la période sur laquelle porte la recherche, à savoir une période où la relance de l'énergie nucléaire a été publiquement débattue et où le gouvernement américain cherchait à reformuler sa politique de gestion des déchets nucléaires. / This thesis approaches valuation as an on-going, constructive and contested process. She addresses this theoretical issue through the study of a very complex object of valuation: nuclear waste in the context of the United States. As an ontologically ambivalent object, nuclear waste can be valued or devalued from many angles, which provides an intriguing and exciting test bed to unfold a sociology of valuation. The thesis examines a multitude of sites where the question of the formulation of nuclear values is being raised. These sites are, for example, the design and vicissitudes of the budgetary process conceived to finance the North American nuclear waste program, the trials set up in order to distribute the financial responsibility of a material expected to remain hazardous during the next million years, the efforts of a group of actors to attach the future of their community to the future of nuclear waste, or the uses of an economic convention to estimate the economic value of spent nuclear fuel. This thesis shows that processes of valuation are never limited to the object that is subjected to valuation and proposes the notion of revaluation, first, to articulate the intertwined relationship between the processes of evaluation and valuation, and second, to signify the particularity of the period during which the research has been undertaken, namely a moment when the revival of nuclear energy was publicly debated, and a moment when the U.S. government was seeking to reformulate its nuclear waste policy.

The thesis titled Legal aspects of constructing a deep geological repository od nuclear waste in the Czech Republic deals with the administrative procedures that needs to be taken before beginning construction of a deep geological repository. Work can be divided into three major parts. The first part deals with analysis of current legislation relating to nuclear energy, with emphasis on the treatment of nuclear waste. International and European commitments of the Czech Republic, current and new Atomic act, as well as institutional and financial arrangements for nuclear waste management are also included in this part. The following section has been devoted to the various administrative procedures. The goal of this section in not to provide complete description of the procedures, but to highlight points of interest and identify potential problems of current legislation and to propose better solution. The last major part is dedicated to public participation in the various administrative procedures. Emphasis is placed on the possibility of involvement of public and the affected communities in related administrative procedures. This chapter also contains a draft of the bill on community involvement in the process of selecting a site for deep repository.

Canada produces about one-third of the global supply of medical radioisotopes. The nuclear power reactors in Ontario, Quebec and New Brunswick have generated about 17 percent of the electricity in the country every year (NWMO, 2010; Noorden; 2013). Since the 1960s, more than 2 million used (or spent) fuel bundles (high-level radioactivity) and 75,000 m³ of low- and intermediate-level radioactive waste have been produced, which is increasing by 2000 to 3000 m³ every year after reducing the processed volume (Jensen et al., 2009). More than 30 countries around the world, including Canada, have proposed construction of very deep geological repositories (DGRs) to store this nuclear waste for design periods 1,000,000 years. DGR concepts under development in Canada (the DGR is likely to be constructed in Ontario) are based on a multi-barrier system (NWMO, 2012). A crucial component of the multi-barrier system is the engineered barrier system (EBS), which includes a buffer, backfill, and tunnel sealing materials to physically, chemically, hydraulically and biologically isolate the nuclear waste. Bentonite-based material has been chosen for this critical use because of its high swelling capacity, low hydraulic conductivity, and for its good ability to retain radionuclides in the case of failed canisters. However, the presence of bentonite-based material in DGRs, surrounded by an aggressive environment of underground saline water, nuclear waste heat decay, and corrosion products under confining stress, may lead to mineralogical changes. Consequently, the physical and physiochemical properties of bentonite-based materials may change, which could influence the performance of bentonite in an EBS as well as the overall safety of DGRs. The objective of this research is to investigate the impact of the underground water salinity, heat generated by nuclear waste, and corrosion products of nuclear waste containers in Ontario on the engineering and micro-/nano-structural properties of bentonite-sand engineered barrier materials. Free-swelling, swelling pressure and hydraulic conductivity tests have been performed on bentonite-sand mixtures subjected to various chemical (groundwater chemistry; corrosion water with iron as a corrosion product) and thermal (heat generated) conditions. Several techniques of micro- and nano-structural analyses, such as x-ray diffraction (XRD), X-Ray microanalysis (DES), surface area and pore size distribution analyses (BET, BJH) and differential gravimetric (TGA and DTG) analyses have also been conducted on the bentonite-sand materials. Valuable results have been obtained for better understanding the durability and performance of the bentonite-sand barrier for the DGR which may be located in Ontario. The obtained results have shown that the groundwater chemistry and corrosion products of the nuclear containers significantly deteriorate the swelling and permeability properties of the tested bentonite-sand barrier materials, while temperature has little or no effect.

The Institute of Resource Ecology (IRE) is one of the eight institutes of the Helmholtz-Zentrum Dresden – Rossendorf (HZDR). The research activities are mainly integrated into the program “Nuclear Waste Management, Safety and Radiation Research (NUSAFE)” of the Helmholtz Association (HGF) and focused on the topics “Safety of Nuclear Waste Disposal” and “Safety Research for Nuclear Reactors”. Additionally, various activities have been started investigating chemical and environmental aspects of processing and recycling of strategic metals, namely rare earth elements. These activities are located in the HGF program “Energy Efficiency, Materials and Resources (EMR)”. Both programs, and therefore all work which is done at IRE, belong to the research sector “Energy” of the HGF. The research objectives are the protection of humans and the environment from hazards caused by pollutants resulting from technical processes that produce energy and raw materials. Treating technology and ecology as a unity is the major scientific challenge in assuring the safety of technical processes and gaining their public acceptance. We investigate the ecological risks exerted by radioactive and nonradioactive metals in the context of nuclear waste disposal, the production of energy in nuclear power plants, and in processes along the value chain of metalliferous raw materials. A common goal is to generate better understanding about the dominating processes essential for metal mobilization and immobilization on the molecular level by using advanced spectroscopic methods. This in turn enables us to assess the macroscopic phenomena, including models, codes, and data for predictive calculations, which determine the transport and distribution of contaminants in the environment.