Les interactions entre le fer métal et l'argilite du Callovo-Oxfordien choisie comme roche hôte pour le stockage des déchets radioactifs, a été étudiée expérimentalement. Le rôle des paramètres clés des transformations (rapports fer/argile, et liquide/roche) ont été étudiés à 90°C, et afin d'accélérer la cinétique des réactions à 150°C et 300°C. Les interstratifiés illite-smectite et les illites sont dissous et remplacés par de nouvelles phases argileuses. Les analyses au MET ont permis de déterminer des Fe-serpentines (phases à 7 Å, groupe de la berthiérine) à 90°C, des phases gonflantes de type smectite tri-octaédrique à 150°C, et des chlorites et des interstratifiés chlorite/smectite à 300°C. Quelle que soit la température, la transformation des feuillets illitiques (I et I/S) n'est pas complète et il reste toujours des feuillets à garniture interfoliaire sodi-calcique, interprétés comme des smectites résiduelles ou néoformées. En comparaison avec le système fer-smectite (bentonite), les différences portent sur la cinétique de réaction, beaucoup plus rapide dans le cas de l'argilite, et l'instabilité du quartz qui comme l'illite contribue à libérer le silicium nécessaire à la formation des silicates de fer. Les transformations observées trouvent des analogies avec les altérations hydrothermales et métasomatoses Fe-Mg des systèmes naturels / The interactions between the iron metal and the argillite from the Callovo-Oxfordian formation chosen as host for the disposal of the radioactive wastes, was experimentally studied. The role of the key parameters of the transformations (iron / clay, and liquid / rock ratios) was studied at 90°C, and, to accelerate reaction kinetics, at 150°C and 300°C. Mixed layered illite-smectite and illites are dissolved and replaced by new clay phases. TEM analyses show that Fe-serpentines (7 Å phases, group of the berthierine) predominates in run products at 90°C, tri-octaedral Fe-rich smectites at 150°C, and chlorites and probably smectite chlorite mixed layered at 300°C. Whatever the temperature, the illite and I/S replacement is not complete and trun products are always accompanied by sodi-calcic residual and/ or newly formed smectites. In comparison with the iron-smectite (bentonite) system, the differences concern the reaction kinetics which are much faster in the case of the argillite, and the instability of the quartz which, as the illite, contributes to release the silicium necessary for the formation of iron silicates. The observed process find analogies with the hydrothermal changes described in natural hydrothermal alterations and Fe-Mg metasomatism

Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2016-04-08T14:08:44Z No. of bitstreams: 0 / Made available in DSpace on 2016-04-08T14:08:44Z (GMT). No. of bitstreams: 0 / Dissertação (Mestrado em Tecnologia Nuclear) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:43:26Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T13:58:17Z (GMT). No. of bitstreams: 1 06498.pdf: 10769439 bytes, checksum: e1653f842e3f8a16356a7f469da93549 (MD5) / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2016-04-08T14:08:44Z No. of bitstreams: 0 / Made available in DSpace on 2016-04-08T14:08:44Z (GMT). No. of bitstreams: 0 / Neste trabalho é apresentado o potencial de remoção de íons estrôncio de soluções aquosas pelas fibras de coco na forma bruta e na forma ativada com peróxido de hidrogênio, 1% e 4%, em meio básico. Os experimentos de biossorção foram realizados em batelada com 2 mg.L-1 de solução de Sr(NO3)2 e foram estudadas as influências dos seguintes parâmetros: tempo de contato, pH e a eficiência de biossorção das fibras ativadas em comparação com a fibra de coco bruta (FCB). A caracterização das fibras antes e após o tratamento, e a presença de Sr2+ nas fibras foram realizadas por microscopia de varredura eletrônica com detector de espectroscopia de energia dispersiva, espectroscopia de difração de raios X, espectroscopia de infravermelho e por análise térmica. Dentre as fibras estudadas, a fibra de coco ativada com 1% H2O2 (FCA 1) apresentou a maior capacidade de biossorção: 3,6 mg.g-1, nas seguintes condições: 5 mg de biomassa em pH 6, após 90 minutos de tempo de contato à temperatura ambiente. A fibra de coco ativada com 4% H2O2 (FCA 2) levou a uma maior degradação dos constituintes da fibra e consequentemente a uma menor remoção de íons de Sr2+.Para os estudos de modelos de isotermas de biossorção de Sr2+, tanto a FCB quanto a FCA 1 ajustaram-se melhor ao modelo de Langmuir e à cinética de pseudo-segunda ordem. Os parâmetros termodinâmicos energia livre de Gibbs (ΔG) e coeficiente de distribuição (KD) foram -0,90 kJ.mol-1 e 265,3L.Kg-1 para a FCB e de -7,2 kJ.mol-1 e 824,3 L.Kg-1 para a FCA1. Esses resultados demonstraram que a FCA 1 possui uma boa eficiência para remover íons de Sr2+de resíduos químicos aquosos e é uma boa alternativa no tratamento de rejeitos radioativos líquidos contendo íons 90Sr. / Dissertação (Mestrado em Tecnologia Nuclear) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

First, the effect of transition metal oxide (e.g., V2O5, Co2O3, etc.) on the physical properties (e.g., density, glass transition temperature (Tg), optical properties and mechanical properties) and chemical durability of a simplified borosilicate nuclear waste glass was investigated. Adding V2O5 in borosilicate nuclear waste glasses decreases the Tg, while increasing the fracture toughness and chemical durability, which benefit the future formulation of nuclear waste glasses. Second, structural study of ZrO2/SiO2 substitution in silicate/borosilicate glasses was systematically conducted by molecular dynamics (MD) simulation and the quantitative structure-property relationships (QSPR) analysis to correlate structural features with measured properties. Third, for bioactive glass formulation, mixed-network former effect of B2O3 and SiO2 on the structure, as well as the physical properties and bioactivity were studied by both experiments and MD simulation. B2O3/SiO2 substitution of 45S5 and 55S5 bioactive glasses increases the glass network connectivity, correlating well with the reduction of bioactivity tested in vitro. Lastly, the effect of optical dopants on the optimum analytical performance on atom probe tomography (APT) analysis of borosilicate glasses was explored. It was found that optical doping could be an effective way to improve data quality for APT analysis with a green laser assisted system, while laser spot size is found to be critical for optimum performance. The combined experimental and simulation approach adopted in this dissertation led to a deeper understanding of complex borosilicate glass structures and structural origins of various properties.

The purpose of the work described in this thesis was to explore the use of fractional crystallization as a technology that can be used to separate medium-curie waste from the Hanford Site tank farms into a high-curie waste stream, which can be sent to a Waste Treatment and Immobilization Plant (WTP), and a low-curie waste stream, which can be sent to Bulk Vitrification. The successful semi-batch crystallization of sodium salts from two single shell tank simulant solutions (SST Early Feed, SST Late Feed) demonstrated that the recovered crystalline product met the purity requirement for exclusion of cesium, sodium recovery in the crystalline product and the requirement on the sulfate-to-sodium molar ratio in the stream to be diverted to the WTP. In this thesis, experimental apparatus, procedures and results are given on scaled-down experiments of SST Early Feed for hot-cell adaptation along with operating parameters and crystallization mechanism studies on early feed multi-solute crystallization. Moreover, guidance is given regarding future steps towards adapting the technology to multi-salt crystallization kinetic parameter estimates and modeling. Crystallization, Evaporative Fractional Crystallization, Nuclear Waste Pretreatment, Cesium Removal, Hanford, SST Early and Late feed, Multi-solute, Multi-salts, Simulant Testing

Submitted by Marco Antonio Oliveira da Silva (maosilva@ipen.br) on 2017-05-25T11:35:08Z No. of bitstreams: 0 / Made available in DSpace on 2017-05-25T11:35:08Z (GMT). No. of bitstreams: 0 / A tecnologia de nanocompósitos é aplicável a uma vasta gama de polímeros termoplásticos e termofixos. A utilização de subprodutos da cana de açúcar tem sido extensivamente estudada como fonte de reforços para os nanocompósitos. O bagaço da cana é largamente utilizado na cogeração de energia e, como resultado da queima deste material, são produzidas milhões de toneladas de cinzas. Para este trabalho, sílica contida nas cinzas do bagaço da cana de açúcar foi extraída por método químico e método térmico. O método térmico se mostrou mais eficiente levando a uma pureza de mais de 93 % em sílica, enquanto o método químico gerou sílica bastante contaminada com cloro e sódio provenientes dos reagentes da extração. As partículas de sílica obtidas foram avaliadas por espalhamento de luz dinâmico (DSL) e apresentaram tamanho médio de 12 μm. Estas partículas foram submetidas à moagem em moinho de bolas e na sequência a tratamento sonoquímico em meio líquido. As partículas de sílica tratadas no processo sonoquímico a 20 kHz, potência de 500 W e 90 minutos tiveram suas dimensões reduzidas a escala nanométrica da ordem de dezenas de nanômetros. A nanossílica obtida foi então incorporada como reforço em polietileno de alta densidade (HDPE). Ensaios mecânicos e termo-mecânicos mostram ganhos de propriedades mecânicas, com exceção da propriedade de resistência ao impacto. O ensaio de deflexão térmica (HDT) mostrou que a incorporação deste reforço no HDPE levou a um pequeno aumento nesta propriedade relação ao HDPE puro. A cristalinidade dos nanocompósitos gerados foi avaliada por meio de calorimetria exploratória diferencial (DSC) e observou-se um decréscimo de cristalinidade do material quando a incorporação de reforço foi de 3%. O material irradiado a 250 kGy com feixe de elétrons mostra ganhos acentuados na principais propriedades do mesmo, principalmente devido ao alto nível de reticulação do HDPE irradiado. / Tese (Doutorado em Tecnologia Nuclear) / IPEN/T / Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP

It is a generally agreed internationally that high level nuclear wastes containing long-lived radioactive wastes should be disposed in deep and stable geological formations that are 500-1000 m below ground level. Deep geological disposal is based on the concept of multiple barriers to prevent deep ground-waters, present in almost all rock formations, from rapidly leaching the wastes and transporting radioactivity away from the repository. The multiple barrier system comprises of ‘engineered barriers’ that are constructed in the repository and ‘natural barriers’ in the surrounding geological environment. The engineered barrier components comprise of the vitrified solid waste, canister (to contain the vitrified waste), and a buffer or backfill material (clay or cement) that fills the annular space between the canister and the walls of the hole drilled in the floor of host-rock. The natural barrier is provided by the rocks and soils between the repository and earth’s surface. The canisters containing the hig level waste (HLW) upon placement in DGR need protection against tectonic activities and chemical attack by dissolved elements and from microbes. Densely compacted bentonite is identified suitable for this purpose owing to its large swell potential, low permeability, sufficient bearing capacity and high cation adsorption capacity. In the deep geological repository (DGR) for disposal of high level nuclear wastes, iodine-129 is one of the significant nuclides, owing to its long half-life (half life = 16 million years) and tendency to easily migrate out of the geological repository into the biosphere caused by its high solubility and poor sorption onto most geologic media. Bentonite buffer by virtue of negatively charged basal surface has negligible affinity for retention of iodide anions. Attempts have been made to improve the iodide retention capacity of bentonite by treating the clay with cationic polymers, this however occurs at the cost of reduced swelling ability of bentonite clay. The compacted bentonite employed in deep geological repositories must possess large swell potential to enable it to close fissures and cracks that form on drying of the expansive clay by the heat arising from the high level nuclear waste and thereby close pathways for migration of radionuclides (from breached canister) to the geo-environment. Therefore, it becomes important to identify an additive that enhances the iodide retention ability of the mix without significantly impairing its swelling ability. Based on the strong affinity of silver for iodide ions, the feasibility of mixing silver-kaolinite (termed AgK) clay with bentonite to improve the latter’s iodide sorption capacity and the impact of mixing AgK clay with bentonite on swelling ability of the mix forms one of the the focus of this thesis. Silver-kaolinite clay was prepared by heating 80% kaolinite + 20% silver nitrate mix at 400°C for 30 min, followed by washing (to remove unreacted silver nitrate) and oven-drying the resultant AgK clay. Physical mixing of AgK and bentonite was considered a viable proposition as small additions (10% to 20% on dry mass basis) besides imparting iodide sorption ability was expected to have minor influence on the swelling ability of the mix. As organo-bentonites are known to retain iodide ions, it was considered relevant to compare the iodide removal behaviour of AgK and organo¬bentonite clay. Hexadecylpyridinium-bentonite (termed as HDPy+B) is the organo¬bentonite examined in this thesis and is prepared by treating bentonite with hexadecylpyridinium chloride mono hydrate salt (C21H38ClN.H2O; molecular weight = 358.01). The hexadecylpyridinium chloride mono hydrate salt is a cationic quaternary ammonium compound and has been used by earlier researchers to prepare organo-bentonite for removal of iodide ions from aqueous solutions. The impact of mixing AgK and HDPy+B clays on the iodide retention and swelling behaviour of bentonite is also considered in the thesis. The mass-balance calculations, XRD analysis, X-ray photon emission survey spectrum and EPMA tests performed on kaolinite-silver nitrate mix/AgK/kaolinite specimen indicated that silver occurs as uniform coatings of AgO/Ag2O on kaolinite surface of the AgK specimen. The AgK clay has strong affinity for iodide ions reflected by the large distribution coefficients (Kd) values of 1367 and 293 mL/g at initial iodide concentrations of 750 mg/L and 1000 mg/L. Further, the sorption process was rapid, unaffected by the presence of co-ions, elevated temperature of sorption and was practically irreversible at range of pH conditions. The iodide retention by AgK is attributed to occurrence of hydrolysis and exchange reactions. On contacting the AgK with water, the AgO species hydrolyze to form AgOH; iodide ions are retained by replacing the hydroxyl group of AgOH leading to formation of AgI phase. The adsorption of HDPy+Cl- ions by bentonite occurs by replacement of the native exchangeable cations by HDPy+ ions and adsorption by van der Waals interactions between the organic cations and the clay surface. The adsorbed cationic polymer neutralize the negative charge of the clay surface. Zeta potential measurements of HDPy+B specimen indicated that adsorption of cationic polymer transforms the negatively charged clay particles into positively charged particles that favour anion adsorption. Sorption of iodide ions by HDPy+B specimen exhibits two distinct segments: 1) the iodide sorption increased rapidly at lower iodide concentration (91 mg/L to 475 mg/L) and are retained by Coulombic adsorption to the cationic groups contained in the loops and tails of the adsorbed polymer (primary adsorption sites) and 2) the relatively slower adsorption at higher iodide concentrations (larger than 475 mg/L) is attributed to exchange with chloride ions attached to HDPy+Cl-ion pair (secondary adsorption sites). The Kd values for iodide adsorption vary from 15 mL/g to 184 mL/g at initial iodide concentrations of 91 mg/L to 996 mg/L respectively. Comparing the iodide removal efficiencies of AgK and HDPy+B specimens revealed that the AgK clay exhibited larger iodide removal; further while the iodide removal by AgK specimen was almost instantaneous (complete in < 5 min), iodide removal by HDPy+B specimen was a slow process (18-24 h is needed to attain equilibrium). Likewise, the iodide retention capacity of the 50%B-50%HDPy+B mix (B = bentonite) is substantially smaller than of the 90%B-10%AgK and 80%B¬20%AgK mixes. Cation exchange capacity (CEC) measurements brought out that mixing AgK with bentonite besides imparting an iodide retention capacity essentially retains the large cation exchange capacity of the expansive clay. On the other hand mixing HDPy+B with bentonite imparts a smaller iodide retention capacity to the mix and leads to a notable reduction in the CEC of the expansive clay. Results of oedometer swell tests brought out that dilution of bentonite with 10% and 20% AgK specimen does not impact its swell potential and leads to some (10%) reduction in swell pressure, while dilution with 50% HDPy+B clay leads to notable (58%) reduction in swell potential and swell pressure (21%) underlining the superiority of AgK specimen as additive to bentonite in deep geological repositories. The swell pressure of the compacted 50%B-50%HDPy+B mix is 21% lower than that of the compacted bentonite specimen. Comparatively, dilution of bentonite with 10% and 20% AgK specimen induces 8-10% lower swell pressure in comparison to the undiluted counterpart. Swell pressure results of compacted 80%B-20%HDPy+B mix is not considered as this mix was unable to retain iodide ions. Superposing the field 129I concentration levels on I removal efficiency indicate that use of 90%B-10%AgK mix would suffice to provide 100% iodide removal efficiency and ensure that the swelling characteristics of bentonite is least affected by dilution.

About 60 percent of the electricity production in Belgium originates from nuclear power plants. Belgium owns 7 nuclear pressurized water reactors, which are located in two sites: 4 reactors in Doel and 3 reactors in Tihange. Together they have a capacity of approximately 5900 MWe. All these reactors use classical uranium oxide fuel assemblies. Two of them (Doel3, Tihange2) have also accepted a limited number of mixed (uranium and plutonium) oxide fuel assemblies. These mixed fuel assemblies came from the reprocessing of spent uranium oxide fuel assemblies in La Hague (France). The reprocessing of spent fuel gives birth to vitrified high-level waste, and to different isotopes of uranium and plutonium, which can be used in the manufacture of mixed oxide fuel assemblies. Each country producing radioactive waste must find a solution to dispose them safely. The internationally accepted solution is to dispose high-level radioactive waste in a deep and stable geological layer. This seems to be the most secure and environment-friendly way to get rid of the high-level radioactive waste. One of the few stable geological layers, which could accept radioactive waste in Belgium, is the Boom clay layer. Another possible layer is the Ypresian clay layer, but it is not the reference option for the moment. The Boom clay layer is quite thin (about 100 m thick) and is not at a large depth (about 240 m below the ground surface) at the proposed disposal site, beneath the SCK CEN Nuclear Research Centre in Mol. A large number of studies have already been performed on the Boom clay layer, and on the possibility of building a high-level radioactive waste repository in this geological medium. Since 1993, the Belgian government has promulgated a moratorium on the reprocessing of spent uranium oxide fuels in La Hague. Since then, spent fuel assemblies are considered as waste, and ONDRAF/NIRAS (the Belgium Agency for Radioactive Waste and Enriched Fissile Materials) has thus to deal with them as waste. This rises a number of questions on how to deal with this new kind of waste. A solution is to directly dispose these spent fuel assemblies in containers in a repository, just like the other high-level radioactive waste. This repository would be build in the Boom clay layer at a depth of about 240 m beneath the SCK CEN. One of the questions raised by this new kind of waste is: "could the direct disposal of the spent nuclear fuel assemblies lead to nuclear criticality risks in the future?". Nuclear criticality is the ability of a system to sustain a nuclear fission chain reaction. This question was not a key issue with vitrified high-level waste because these do not include fissile uranium and plutonium isotopes, which could lead to a criticality event. The spent fuel repository will be designed in order to totally avoid the occurrence of a criticality event at the closure time. But in the future history of the repository, external events could possibly affect this. These events could maybe lead to criticality inside the repository, and this has also to be avoided. This work tries to answer this question, and to determine how to avoid a long-term criticality event inside the repository. The only complete research work answering this question has been performed in the U.S. for the Yucca Mountain repository but this design is fully different from the Belgian one studied here: for example, the waste are not only spent fuel waste, and the geological layer is volcanic tuff.<p>The main achievements of this work are: <p>*A first set of in-package criticality scenarios for different design options for a Belgian spent fuel repository in the Boom clay layer. <p>*A large number of criticality calculations with different parameters (fuel type, fuel burnup, fuel enrichment, distance between the fuel assemblies, distance between the fuel rods, water fraction inside the overpack) for the different design options. <p>*A preliminary study of the effects of the spent fuel assemblies isotopic evolution with time on the multiplication factor. <p>*For the first time, a coupling between the in-package criticality scenarios and the criticality calculations has been performed. / Doctorat en sciences appliquées / info:eu-repo/semantics/nonPublished

Tous les États sans discrimination ont un droit inaliénable de développer les utilisations de l'énergie nucléaire à des fins civiles, à condition de ne pas détourner ces utilisations pacifiques vers des armes nucléaires. Cependant, il est accordé à cinq pays le droit de posséder ces armes, à savoir les États-Unis, la France, la Russie, la Chine et le Royaume-Uni. Autour de cette position, un vif débat à la fois juridique et éthique a été soulevé. En effet, pour ses opposants, le nucléaire représente un risque durable et non maîtrisable par la science. Les accidents nucléaires majeurs, les déchets radioactifs et le détournement du nucléaire à des fins militaires sont des risques ingérables et d‟une gravité exceptionnelle. En revanche, les défenseurs de cette énergie la présentent comme sûre, voire partie prenante du développement durable. Selon eux, le nucléaire est un moyen fiable de lutter contre le réchauffement climatique et aussi une solution à la pénurie énergétique à laquelle le monde est confronté. En examinant et analysant la fiabilité et la crédibilité de tous les arguments allant à l‟encontre et en faveur de cette industrie, on constate que la licéité et la légitimité du recours à l'énergie nucléaire sont mal fondées. Par conséquent, nous estimons qu‟il est nécessaire de dépasser le nucléaire par la conclusion d'une convention internationale posant l'interdiction progressive mais complète du nucléaire. / All states without discrimination have an inalienable right to develop the uses of nuclear energy for civilian purposes, provided they do not divert these peaceful uses to nuclear weapons. However, five states have been granted the right to possess these weapons, that is : United-States, France, Russia, China and United-Kingdom. Around this position a fierce debate, both legal and ethical, has been raised. Indeed for its opponents nuclear represents a persistent risk that is non controllable by science. Major nuclear accidents, radioactive wastes and the use of nuclear for military purposes are unmanageable risks of exceptionnal serious gravity. On the other hand, the proponents of this energy present it as safe, even as part of sustainable development. According to them, nuclear is a reliable means to fight global warming and is also a solution to the energy shortage the world is facing. When analyzing the reliability and the credibility of all arguments for and against this industry, it can be noticed that the lawfulness and legitimacy of the use of nuclear energy are ill-founded. Therefore, we believe there is a need to go beyond nuclear with the conclusion of an international convention dealing with the progressive but comprehensive nuclear ban.