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

Thesis: M.S., Massachusetts Institute of Technology, Department of Nuclear Engineering, 1977 / Includes bibliographical references. / M.S. / M.S. Massachusetts Institute of Technology, Department of Nuclear Engineering

This study includes a systematic investigation of the bond strength of cementitious borehole plugs in welded tuff. Analytical and numerical analysis of borehole plug-rock stress transfer mechanics is performed. The interface strength and deformation are studied as a function of Young's modulus ratio of plug and rock, plug length and rock cylinder outside-to-inside radius ratio. The tensile stresses in and near an axially loaded plug are analyzed. The frictional interface strength of an axially loaded borehole plug, the effect of axial stress and lateral external stress, and thermal effects are also analyzed. Implications for plug design are discussed. Push-out tests are used to determine the bond strength by applying an axial load to the cement plugs. A total of 130 push-out tests are performed as a function of borehole size, plug length, temperature, and degree of saturation of the tuff cylinder. The use of four different borehole radii enables evaluation of size effects. A well-defined exponential strength decrease with increasing plug diameter results.

This study is about the impacts of the transportation of low level radioactive waste (LLRW) on American Indians. The terms American Indians, Native Americans, and Indians are used interchangeably throughout this report to refer to people who are members of tribes in the United States. The information contained in this report is valuable to non -Indian individuals, communities, and governments as well as to the tribes and the U.S. Department of Energy/Nevada Operations Office (DOE/NV) for which it was prepared. Many of the individuals who agreed to participate in this study asked if their non -Indian neighbors were also being given the opportunity to share their views and perspectives on the transportation of LLRW near and through their neighborhoods. Although this study was designed to include only Native Americans, it can serve as a model for additional studies in non –Indian communities. American Indian tribes have a unique status as sovereign nations within the U.S., and this study was designed to address that relationship.This study includes an assessment of social and cultural impacts. One type of impact assessment concerns the estimation and communication of risks associated with potentially dangerous technologies or substances. Such an assessment, a technological "risk assessment," is generally conducted by natural or physical scientists and focuses on the probability and magnitude of various scenarios through time (Wolfe 1988). The specialists who conduct the assessment believe their estimates reflect the "real risks" of a technology or project because the estimates were made using scientific calculations. This study is not a risk assessment. Instead, this study pays attention to the public perceptions of impacts and risks. Like other social scientists, the researchers and American Indian partners who designed and conducted this study focus on public perceptions and frame the discussions in terms of locally defined values and concerns.This study involves 29 tribes and subgroups and is therefore very complex. Every effort has been made to present information systematically to help the reader make sense of what is being presented. Information about the tribes is presented in the same order throughout the report.

The topic of this diploma thesis is Czech Nuclear Law. The nuclear law is a set of legal norms concerned with the ways and conditions of usage of nuclear energy and ionising radiation, the systems used to protect people and the environment against the damaging effects of ionising radiation, protocols to be followed in the case of accidents, radioactive waste disposal as well as the role of the state administration in supervising usage of nuclear energy and ionising radiation. A short historical review of the usage of nuclear energy and ionising radiation is followed by 2 chapters dedicated to international and EU nuclear law. The aims and content of international and EU collaboration concerning non-military usage of nuclear energy and ionising radiation are defined in these chapters. The other chapters of this thesis are concerned with Czech Nuclear Law. The constitutional foundation of this Law is described as well as nuclear law as a legal subject and the rules forming the base of Czech Nuclear Law. The following section deals with the power characteristics of the individual state authorities in the field of non-military usage of nuclear energy. Attention is paid to the classification of ionising radiation sources and characteristics of the tools used by the legal norm for environmental...

Rejeitos radioativos são gerados em todas as etapas do ciclo do combustível nuclear e em instituições que fazem uso de elementos radioativos. No Brasil, para que possam atender a legislação vigente, eles devem ser devidamente gerenciados e tratados. A caracterização é a etapa do gerenciamento de rejeitos radioativos que certifica e complementa as informações presentes no documento de identificação fornecido pelo gerador dos rejeitos radioativos. O objetivo desse trabalho foi desenvolver uma metodologia para caracterização de tambores de rejeitos radioativos. Neste trabalho foi utilizada a técnica de espectrometria gama associada ao método de Monte Carlo para obter e estabelecer uma relação entre as eficiências de contagens teóricas e experimentais para quatro tambores de calibração com densidades diferentes, preparados com nove tubos de PVC em seu interior. O detector de HPGe utilizado foi posicionado em três diferentes distâncias em relação a extremidade do tambor de calibração. A utilização do método de Monte Carlo mostrou-se adequado a metodologia proposta neste trabalho. / Radioactive wastes are generated at all stages of the nuclear fuel cycle and institutions that make use of radioactive elements. In Brazil, so that they can meet the current legislation, they must be properly managed and treated. The characterization is the stage of radioactive waste management that certifies and complements the information in the identification document provided by the radioactive waste generator. The aim of this study was to develop a methodology for characterization of radioactive waste drums. In this work we used the gamma spectrometry technique associated with the Monte Carlo method to obtain and establish a relationship between the efficiencies of theoretical and experimental scores for four calibration drums with different densities prepared with nine PVC tubes inside. The HPGe detector used was positioned at three different distances from the end of the calibration drum. Using the Monte Carlo method was appropriate to the methodology proposed in this work.

Todo e qualquer tipo de rejeito deve ser tratado e armazenado adequadamente. Portanto os rejeitos radioativos requerem gerenciamento apropriado e seguro, desde sua geração até seu armazenamento em repositório. O principal propósito da gerência de rejeitos radioativos é preservar a vida humana e o meio ambiente. O objetivo da pesquisa foi caracterizar algumas argilas brasileiras de modo a avaliar a viabilidade de seu uso na camada de recheio, uma das barreiras de um repositório de rejeitos radioativos. A principal função desta barreira é contribuir para retardar o movimento de radionuclídeos e previnir sua liberação para o ambiente. Quatro argilas de fornecedores nacionais foram selecionadas para a pesquisa: montmorilonita cálcica (Dol 01), montmorilonita sódica (Dol 02), caulinita (Ind 01) e vermiculita (Ubm 04). Foram determinadas suas caracteristicas físicas, químicas e mineralógicas e também seu potencial de sorção dos cátions césio e o estrôncio. Foi confirmada por meio destes resultados relação direta entre a superfície específica (SE), a capacidade de troca catiônica (CTC) e o pH destas argilas, na seguinte ordem crescente: Ind 01, Dol 01 e Dol 02. De acordo com os modelos de Freundlicdh (Kf) e Langmuir (M), as argilas Dol 01 e Dol 02 foram melhores sorvedoras de Sr2+. A variação de energia livre de Gibbs também foi calculada para as reações de sorção entre as argilas e os cátions e para todas as argilas, esta variação for negativa, confirmando a espontaneidade das reações de sorção. / Wastes in general should be properly treated and stored. Then the radioactive wastes also rquire suitable and safe management beginning in their generation until the storage in repository. The main purpose of the radioactive waste management is to preserve the human beings and the environment. The objective of this research ws to characterize some brasilian clays in order to evaluate of their use in the backfill layer, one of the radioactive waste repository barriers. The main function of this barrier to contribute in the delay of the radionuclides movement, and to prevent their release into the environment. Four clays provided by national suppliers were selected for the research: Ca-Montmorillonite (Dol 01), Na-Montmorillonite (Dol 02), Kaolinite (Ind 01) and Vermiculite (Ubm 04). Their physical, chemical and mineralogical characteristics were determined, and also their sorption potential of Cesium and Strontium cations. It was confirmed through these results a direct relationship among their specific surface (SS), the capacity of cationic exchange (CCE) and pH. The CCE results followed this increasing order: Ind 01, Dol 01, and Dol 02. In accordance with the models of Freundlich (Kf) and Langmuir (M), the clays Dol 01 and Dol 02 were the best sobers of Sr2+. The Ind 01 and Ubm 04 were the best ones in the case of Cs+. The Gibbs free energy change was calculated for the sorption reactions between the clays and the cations, and it was negative for all clays, confirming the sorption spontaneity.

O reprocessamento do elemento combustível constitui uma das etapas do ciclo do combustível e tem por objetivo separar U 235, U238 e Pu239 dos produtos de fissão contidos no elemento combustível queimado para reutilizá -los na fabricação de novos elementos combustíveis. No Brasil, o desenvolvimento do reprocessamento teve início nos anos 70 com a implantação de uma planta piloto de reprocessamento, no Instituto de Pesquisas Energéticas e Nucleares Comissão Nacional de Energia Nuclear/São Paulo (I PEN-CNEN/SP), dando origem à uma instalação, em escala de laboratório, que operou até o início da década de 90. Parte dos rejeitos radioativos, produzidos principalmente a partir dos laboratórios analíticos está armazenados à espera de tratamento no Labora tório de Gestão de Rejeitos do IPEN-CNEN/SP, e são constituído pela mistura de fases aquosa e orgânica. Por conter compostos orgânicos, esses rejeitos não podem se solidificados diretamente com cimento, necessitando para isso, de tratamento prévio para torná-los compatíveis. Desta forma, o objetivo deste trabalho foi desenvolver uma metodologia de biodegradação dos compostos orgânicos que compõem os rejeitos radioativos líquidos orgânicos (RRLO) para que os mesmos possam ser posteriormente imobilizados em c imento. O trabalho foi dividido em três etapas, a de caracterização dos RRLO, obtenção das comunidades microbianas (da mina de urânio de Poços de Caldas e do estuário de São Sebastião) e realização de ensaios de biodegradação dos RRLO. A partir da caracterização dos rejeitos, a biodegradação do TBP e acetato de etila foram monitoradas. Os resultados mostraram que as comunidades bacterianas selecionadas foram eficientes para a degradação dos rejeitos radioativos líquidos orgânicos. Ao final dos experimentos os níveis biodegradação foram de 66% para o acetato de etila e 70% para o TBP. / The research and development program in reproces sing of low burn-up spent fuel elements began in Brazil in 70s, originating the lab -scale hot cell, known as CELESTE located at Nuclear and Energy Research Institute, IPEN - CNEN/SP. The program was ended at the beginning of 90s, and the laboratory was c losed down. Part of the radioactive waste generated mainly from the analytical laboratories is stored waiting for treatment at the Waste Management Laboratory, and it is constituted by mixture of aqueous and organic phases. The most widely used technique for the treatment of radioactive liquid wastes is the solidification in cement matrix, due to the low processing costs and compatibility with a wide variety of wastes. However, organics are generally incompatible with cement, interfering with the hydration and setting processes, and requiring pre -treatment with special additives to stabilize or destroy them. The objective of this work can be divided in three parts: i) organic compounds characterization in the radioactive liquid waste; ii) the occurrence of b acterial consortia from Poços de Caldas uranium mine soil and São Sebastião estuary sediments that are able to degrade organic compounds and third, the development of a methodology to biodegrade organic compounds from the radioactive liquid waste aiming th e cementation From the characterization analysis, TBP and ethyl acetate were chosen to be degrade d. The results showed that selected bacterial consortia were efficient for the organic liquid wastes degradation. At the end of the experiments the biodegradat ion level were 66% for ethyl acetate and 70% for the TBP.

O uso e a aplicação cada vez mais constante da tecnologia nuclear consistente em áreas relacionadas à saúde, energia, industrial, bélica, agrícola, entre outras, faz com que haja a necessidade de uma regulamentação de acordo com os padrões de segurança e proteção radiológica internacionais. Dessa forma, utilizando-se de conceitos provenientes do Direito Constitucional, do Direito Ambiental e do Direito do Trabalho, o enfoque da presente pesquisa foi investigar a difícil questão da competência nuclear e a competência ambiental, a impossibilidade de legislar dos Estados, bem como a falta de regulamentação sobre Rejeitos radioativos. Para tanto, foram atualizados e revisados critérios e métodos de interpretação constitucional para solucionar possíveis antinomias jurídicas advindas de múltipla positivação de normas pelos entes federados que dificultam tanto o asseguramento quanto o aprimoramento da Proteção radiológica do trabalhador e do meio ambiente. Finalmente, a hipótese considerada demonstrou que as mudanças na estrutura legislativa nas três esferas de poderes são necessárias, visando à aplicabilidade de responsabilidade legal na esfera nuclear, principalmente no que se refere às entidades administrativas e estatais. / The use and increasingly steady application of consistent nuclear technology in areas related to health, energy, industrial, war, agriculture, among others, means that there is a need for regulation in accordance with the safety standards and international radiological protection . Thus, using concepts from the Constitutional Law, Environmental Law and Labor Law, the focus of this research was to investigate the difficult issue of nuclear competence and environmental responsibility, the impossibility of legislating states, as well as the lack of regulation on radioactive waste. Therefore, it has been updated and revised criteria and methods of constitutional interpretation to solve possible legal antinomies arising from multiple positivation of the federal entities rules that hinder both the assurance and the improvement of radiation protection of workers and the environment. Finally, the working hypothesis has shown that changes in the legislative framework in the three spheres of power is needed in order to applicability of legal liability in the nuclear sphere, especially with regard to administrative and state entities

Dans le but de vérifier l'efficacité des dispositifs de scellement ou des barrières ouvragées dans le stockage géologique des déchets radioactifs, l'Institut de Radioprotection et de Sûreté Nucléaire (IRSN) a mis en œuvre le projet expérimental SEALEX (SEALing performance EXperiments) auquel ce travail est étroitement lié. Dans le cadre de ce projet, des essais in-situ sont effectués à l'échelle représentative et dans des conditions naturelles sur un mélange compacté de bentonite et de sable. Ce matériau de mélange a été choisi pour sa faible perméabilité et surtout pour sa capacité de gonflement qui permet de colmater les vides existant dans le système, notamment le vide technologique correspondant au vide radial entre le noyau de scellement et la roche hôte et qui est inévitable au cours de l'installation du noyau dans le forage. Une fois les vides scellés, le gonflement à volume constant engendre une pression de gonflement aussi bien sur la roche hôte (radiale) que sur les structures de confinement en béton (axiale). Le comportement de ce matériau dans ces conditions de couplages hydromécaniques est alors étudié dans ce travail. La microstructure du matériau à son état initial a été premièrement examinée par micro-tomographie rayons-X. Ceci a permis de voir la distribution des grains de bentonite et de sable ainsi que le réseau de pores dans l'échantillon. Des macro-pores se sont retrouvés concentrés à la périphérie de l'échantillon ainsi qu'entre les grains de sable, ce qui pourra affecter à court terme la perméabilité. L'hydratation du même matériau en condition de gonflement limité a été ensuite observée par une photographie 2D et par la micro-tomographie aux rayons-X. Le mécanisme de gonflement par production de gel de bentonite, la cinétique de gonflement, la diminution de densité et l'homogénéisation du matériau final on été analysés. L'hydratation en conditions de gonflement empêché a été aussi étudiée par des essais où la pression de gonflement a été mesurée dans deux directions : radialement et axialement. La différence retrouvée entre les pressions de gonflement axiales et radiales a évoqué la présence d'une anisotropie de microstructure qui a été analysée en fonction de la masse volumique sèche de bentonite dans le mélange. Des essais en modèle réduit reproduisant à une échelle 1/10ème les essais in situ (SEALEX) ont été également effectués afin d'étudier le comportement du noyau compacté après la reprise des vides au cas d'un accident détruisant les éléments de confinement. Des mesures locales de pression de gonflement le long des échantillons ont permis de mettre en évidence l'évolution du gradient de densité durant le gonflement axial. Finalement une comparaison entre les résultats obtenus dans ce travail et ceux d'un essai in situ (SEALEX) a été faite. Une bonne correspondance entre les valeurs d'humidités relatives a été retrouvée pour les mêmes longueurs d'hydratation tout en prenant en compte la saturation par le vide technologique radial. Par contre, la comparaison des évolutions et des valeurs de pressions de gonflement était plus compliquée vu les différences de configurations des essais / In order to verify the effectiveness of the geological high-level radioactive waste disposal, the French Institution of Radiation protection and Nuclear Safety (IRSN) has implemented the SEALEX project to control the long-term performance of swelling clay-based sealing systems, and to which this work is closely related. Within this project, In-situ tests are carried out on compacted bentonite-sand mixture in natural conditions and in a representative scale. This material is one of the most appropriate sealing materials because of its low permeability and good swelling capacity. Once installed, this material will be hydrated by water from the host-rock and start swelling to close all gaps in the system, in particular the internal pores, rock fractures and technological voids. Afterwards, swelling pressure will develop. In the present work, laboratory experiments were performed to investigate the sealing properties under this complex hydro-mechanical conditions taking into consideration the effect of technological voids. The microstructure of the material in its initial state was first examined by microfocus X-ray computed tomography (µCT). This allowed identification of the distribution of grains of sand and bentonite as well as the pores in the sample. Macro-pores are found concentrated at the periphery of the sample and between the grains of sand, which could affect in the short term the permeability. The hydration of the same material in limited swelling conditions was then observed by 2D photography and 3D µCT. The swelling mechanism with bentonite gel production, the swelling kinetics, the density decrease and the homogenisation of the material were analyzed. The hydration in the conditions of prevented swelling was also studied by swelling pressure tests with radial and axial measurements of swelling pressure. The difference found between the axial and radial swelling pressures suggested the presence of an anisotropic microstructure. Mock-up tests at a 1/10 scale of the in situ SEALEX tests were carried out for the study of the recovery capacity of the mixture in case of an accident causing the failure of the confining structures. Local measurements of swelling pressures along the sample allowed analysis of the density gradient evolution during axial swelling. Finally, a comparison between the laboratory results and those from an in-situ test was done, showing a good fitting in the relative humidity curves for the same infiltration length while considering the saturation effect from the technological void. The swelling pressure comparison was more complex because of the different configurations of the tests (existence of technological void in-situ that could affect the kinetics)