Made available in DSpace on 2014-10-09T12:51:03Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:07:06Z (GMT). No. of bitstreams: 1 11130.pdf: 14304463 bytes, checksum: 273c21b7633910d5e722f76e468344e2 (MD5) / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Made available in DSpace on 2014-10-09T12:51:56Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:07:55Z (GMT). No. of bitstreams: 0 / Atualmente são muito discutidos temas que abordam a preservação do meio ambiente, para o desenvolvimento de tecnologias de produção que não a agridam, gerando resíduos menos tóxicos e em menor quantidade. Resíduos poluentes contendo metais como o crômio, têm sido lançados nos solos e rios, degradando a água utilizada para o consumo humano. Não diferentes são os problemas decorrentes de atividades nucleares, as quais geram rejeitos nas instalações e laboratórios de pesquisa. Embora estes rejeitos não sejam lançados no meio ambiente, muitas vezes encontram-se armazenados em laboratório inadequadamente, o que pode resultar em graves acidentes. Na intenção de solucionar estes problemas, existem várias técnicas para o tratamento de rejeitos, entre elas a troca iônica eletroquímica (EIX – electrochemical ion exchange). A EIX é um processo avançado que une as vantagens da troca iônica convencional com o fato de usar como reagente o elétron, reduzindo consideravelmente o volume da solução a ser tratada. Esta técnica consiste na elaboração de um eletrodo, no qual o trocador iônico é incorporado fisicamente em uma estrutura do eletrodo com um aglutinante. Optou-se neste trabalho pela resina catiônica Amberlite CG-50 para o tratamento dos rejeitos contendo íons crômio e o trocador catiônico inorgânico fosfato de zircônio para os íons césio, pois apresentam boa estabilidade química em meio oxidante e perante radiação ionizante. A quantidade de carvão, de grafita e aglutinante para a formulação do eletrodo mais eficiente também foi estudada. Após a escolha dos melhores eletrodos, verificaram-se retenções para o Cr e para o Cs da ordem de 99,3% e 99,8%, respectivamente. A eluição completa tanto do íon crômio quanto do íon césio, sem nenhuma adição de reagentes, revelou-se uma das principais vantagens deste processo, o que torna possível a reutilização do eletrodo sem perda de sua capacidade. Com base nos resultados apresentou-se um processo contínuo de tratamento de rejeitos utilizando-se uma célula eletrolítica de fluxo (CELFLUX) de alta capacidade de retenção para o íon Cr e Cs. A alta eficiência desta célula tanto na retenção quanto na eluição, levando a uma redução importante no volume do rejeito e, até mesmo, possibilitando a reutilização dos íons separados, torna o processo altamente viável para o emprego industrial. / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Submitted by Maria Eneide de Souza Araujo (mearaujo@ipen.br) on 2015-02-24T19:40:19Z No. of bitstreams: 0 / Made available in DSpace on 2015-02-24T19:40:19Z (GMT). No. of bitstreams: 0 / Rejeitos radioativos líquidos contendo compostos orgânicos precisam de atenção especial, porque os processos de tratamento disponíveis são caros e difíceis de serem gerenciados. A biossorção é uma potencial técnica de tratamento que tem sido estudada em rejeitos simulados. O termo biossorção é utilizado para descrever a remoção de metais, metalóides e/ou radionuclídeos por um material de origem biológica independentemente de sua atividade metabólica. Dentre as potenciais biomassas, os resíduos agrícolas apresentam características muito atraentes, pois possibilitam a remoção dos radionuclídeos presentes no rejeito utilizando um biossorvente de baixo custo. O objetivo deste estudo foi avaliar o uso potencial de diferentes biomassas originadas de produtos ou resíduos agrícolas (fibra de coco, casca de café e casca de arroz) no tratamento de rejeitos radioativos líquidos orgânicos reais. Foram realizados experimentos com essas biomassas incluindo i) Preparação, ativação e caracterização das biomassas; ii) Realização dos ensaios de biossorção e iii) Avaliação do produto da imobilização de biomassas em cimento. As biomassas foram testadas nas formas brutas e ativadas. A ativação foi realizada com soluções diluídas de HNO3 e NaOH. Os ensaios de biossorção foram realizados em frascos de polietileno, nos quais foram adicionados 10 mL do rejeito radioativo ou diluições do rejeito em água deionizada com o mesmo pH e 2 % da biomassa (m/v). No final do experimento, a biomassa foi separada por filtração e a concentração dos radioisótopos remanescente no filtrado foi determinada por ICP-OES e espectrometria gama. O rejeito estudado contém urânio natural (U (total)), amerício-241 e césio-137. Os tempos de contato adotados foram 30 min, 1, 2 e 4 horas e as concentrações estudadas variaram entre 10% e 100%. Os resultados foram avaliados por meio da capacidade máxima de sorção experimental e modelos ternários de isotermas e cinética. As maiores capacidades de sorção foram observadas com casca de café bruta, com valores aproximados de 2 mg/g de U (total), 40 x10-6 mg/g de Am-241 e 50 x10-9 mg/g de Cs-137 e, também, com fibra de coco ativada, com valores de 2 mg/g de U (total), 70 x10-6 mg/g de Am-241 e 40 x10-9 mg/g de Cs-137. As propriedades avaliadas na determinação da qualidade do produto de imobilização foram água livre, resistência mecânica, trabalhabilidade e tempo de pega. Os melhores produtos de imobilização para estas biomassas foram obtidos com uma relação água/cimento de 0,30, contendo 5%, 10% e 15% de casca café bruto, fibra de coco ativada e casca de café ativado, respectivamente. Estes resultados sugerem que a biossorção com casca de café bruta e fibra de coco sob a forma ativada podem ser aplicadas no tratamento de rejeitos radioativos líquidos orgânicos contendo urânio, amerício-241 e césio-137. / Tese (Doutorado em Tecnologia Nuclear) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2015-12-10T16:58:20Z No. of bitstreams: 0 / Made available in DSpace on 2015-12-10T16:58:20Z (GMT). No. of bitstreams: 0 / O descarte de resíduos químicos é uma prática lesiva ao meio ambiente e à saúde da população. Nesse trabalho foi realizado levantamento dos impactos socioambientais com base nos registros de descartes de resíduos químicos da Companhia Ambiental do Estado de São Paulo - CETESB, no período de 2005 a 2013, tomando como referência a Região Metropolitana de São Paulo - RMSP, em especial as cidades de São Paulo e Guarulhos. Dentre os resultados que nortearam o trabalho se destacam a caracterização da prática de descarte de resíduos químicos, sobrepondo a mapas temáticos georreferenciados de rodovias e recursos hídricos. Também por meio de pesquisa dirigida aos principais atores intervenientes dos cenários acidentais,foi possível conhecer os processos atuais de intervenção e de tratamento, aplicados nos descartes de resíduos químicos nas esferas municipais e estaduais. Nesse contexto, o Decreto Estadual nº 59.263/2013, que regulamenta a Lei nº 13.577/2009 sobre a proteção da qualidade do solo e gerenciamento de áreas contaminadas criou o Fundo Estadual para Prevenção e Remediação de Áreas Contaminadas (FEPRAC), destinado à identificação e remediação das áreas órfãs. O FEPRAC apresentase como um instrumento econômico capaz de elidir o perigo nos casos de emergências químicas envolvendo o descarte de resíduos sem a identificação do responsável. / Tese (Doutorado em Tecnologia Nuclear) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Of the various dissolved species contained in the effluents from the mining and milling of uranium ores, the one which is of particular concern for environmental protection is Radium-226. The literature shows that, in recent years, considerable efforts have been made to develop treatment systems that can achieve the stricter effluent discharge standards imposed by the regulatory bodies. There has also been a concern to treat the already existent sludges from previous treatments. The main priority is to limit, as much as possible, the arising of sludge from future treatment systems. The most common treatment used is the addition of lime and limestone to raise the pH followed by barium chloride to form a very finely divided Ba(Ra)S04 precipitate which is then settled in large ponds or basins. In spite of the high decontamination factors obtained with this technique, these may not be satisfactory in terms of environmental protection. In addition, the industry is increasingly aware of the economical benefits resulting from treatment processes that allow water reuse to the process.

This work consists in development of expert system intended for choosing appropriate method of waste processing. The software is created in VisiRule software which is built on Prolog language and is part of WinProlog 4.900 development tool. It also deals with problematics of creating of knowledge base for applications of this type and judging of suitability of possible approaches to creating an expert system for given purpose.

<p dir="ltr">Risk assessment involves analyzing potential accident scenarios to identify hazards and assess associated risk factors. Nuclear safety and security share the common goal of protecting against radiation exposure. However, they have developed separately, each with their own distinct risk assessment methodologies. As a result, there is a need for a comprehensive risk assessment method that covers both safety and security aspects. This thesis proposes a methodology that integrates risk assessment approaches for nuclear safety and security to address the gap in the current development of their risk assessment methodologies.</p><p dir="ltr">The proposed methodology applies the existing probabilistic risk assessment (PRA) methodology to the PFRI (Potential Facility Risk Index), originally developed to evaluate quantitative nuclear security risks, to enable the inclusion of safety risks in the assessment. The PFRI framework and methodology are modified to ensure that the PFRI score accurately reflects the overall risk of the facility for both nuclear safety and security. The facility-based approach of the existing PFRI is maintained to ensure a comprehensive assessment of the research reactor.</p><p dir="ltr">To achieve the goal of developing a comprehensive risk assessment method, the traditional PRA tools, such as event tree analysis (ETA) and fault tree analysis (FTA), are utilized in combination with the modified PFRI methodology. In addition, the consequence analysis method of PFRI is changed using the MACCS, which is commonly used for consequence analysis in PRA. The modified methodology is then used to conduct a risk assessment for the PFRI by setting safety and security scenarios at a hypothetical nuclear facility. The results demonstrate that the modified PFRI can provide a reasonable traditional risk unit and enable the comparison of risks from both safety and security aspects.</p><p dir="ltr">The final goal of this study is to develop the PFRI to determine the overall risk of the facility, considering both nuclear safety and security aspects. The PFRI score is utilized as a quantitative measure to show the total risk associated with hypothetical nuclear facility, providing a comprehensive understanding of its safety and security. By developing a methodology that integrates risk assessment approaches for nuclear safety and security, this thesis contributes to the improvement of the risk assessment methodology for nuclear facilities.</p>

<p dir="ltr">Anomaly detection systems identify abnormal behaviors, and can increase the uptime, safety, and profitability of an industrial system. This research investigates the development of an AI model for detecting anomalies in nuclear reactors. An LSTM network was used to predict the value of a key reactor signal, and then the predictions are compared to the measured values in order to determine if the data is abnormal. The predictive AI model was trained using regular operation data from the nuclear reactor at Purdue University, PUR-1. It is shown in the experiment that the model can accurately track reactor neutron counts during normal operation, with an average error of less than 5% when predicting five seconds into the future. It is also shown that the model reacts to abnormal inputs, with average errors above 50% when fed data which simulates a false data injection cyberattack. The framework of using prediction error to identify anomalies is investigated and a false positive rate of 0.2% is achieved on the normal evaluation dataset while still identifying the abnormal data as anomalous.</p>

<p dir="ltr">Nuclear mobile microreactors will serve as a unique, transportable power source, especially for remote communities. Because mobile microreactors are meant to remain operable after initial startup, keeping the microreactor cool during transport is a safety concern that must be taken into consideration. Due to the compact nature of shipping containers, there is no significant room for the installation of active cooling mechanisms. The thermal limitations imposed by current regulatory guidelines lead to a reactor shipment external maximum temperature of 85◦C. Transporting heat from the microreactor's exterior walls to its surrounding air within the shipping container under natural convection will serve as the greatest source of resistance to the decay heat removal. In the event of mechanical failure or local regulations restricting forced cooling systems within the shipping container, natural convection will be the primary method for transferring heat. Before mobile microreactors can reach commercial status, research must be conducted on ensuring continued passive safety. </p><p><br></p><p dir="ltr">During the unavailability of helium circulation, the internal reactor core is designed to cool by block-to-block conduction and radiation, and the reactor vessel surface is cooled by the ambient air. This scenario is anticipated during the transport of the micro-high temperature gas-cooled reactor (HTGR) in a shipping container. The conduction and radiation between the prismatic micro-HTGR blocks in the core can be influenced by variances in the thermal contacts. This work investigated the conduction within a simulated horizontal micro-HTGR core. An experimental setup was used to validate a numerical model for conduction radiation cooldown with postulated thermal contact conductances (TCC). The experimental setup consisted of a hexagonal assembly with scaled prismatic blocks placed within a high-temperature vacuum environment. The gaps between the blocks were well controlled and monitored. The experimental setup was designed in such a way that the temperature variation in the axial direction was minimal, such that the experiment could be observed as a 2D (r,θ) heat transfer problem. The experimental scenario was computationally modelled with a finite element analysis (FEA) program. Once validated, the computational model was used to investigate the impact of gap conductance on overall decay heat removal. Using a conservative estimate for gap conductance value (100 W/m2 − K) between the prismatic blocks, there is a negligible increase in temperature observed during decay heat generation with constant natural convection coefficients. </p><p><br></p><p dir="ltr">However, the internal temperature profile may change drastically depending on the exterior conditions of the microreactor. A second model for the worst case scenario of exterior cooling being limited to natural convection flows was validated against existing benchmark experimental data on natural convection in closed cavities. The investigations have been performed for several configurations, including different reactor sizes, power levels, and scenarios with or without shielding around the reactor pressure vessel (RPV). This conservative safety analysis restricts the power level of the reactor to be equal to 1 MWe. A more realistic analysis with intermittent shutdown of shipping container air circulation demonstrates that a 4 MWe reactor will reach 85◦C Code of Federal Regulations (CFR) limitations after one hour while a 5 MWe reactor reaches the limit after 34 minutes. </p><p><br></p><p dir="ltr">Finally, both models were combined into a conjugate heat transfer model to examine whether thermal contact conductance (TCC) values would affect the external temperature profile as well as the maximum temperature reached by the core to ensure material limitations would be maintained. Studies have been conducted on a micro-HTGR design that originates from the fuel block design of the MHTGR-350 with changes to the overall power level, TCC values, and outer shipping container wall temperatures. Changes to TCC values do not significantly change microreactor exterior temperatures. In addition, the internal temperatures under all examined conditions remained under 875◦C. </p>

<p dir="ltr">In the last few decades, the US Department of Energy established the Generation-IV Initiative to advance the design of nuclear energy systems with a focus on fast nuclear reactors. Special interest has been placed on Sodium-cooled Fast Reactors (SFRs) along with metallic fuels. SFRs are attractive because of their ability to utilize fast neutrons effectively, allowing for efficient transmutation of long-lived radioactive isotopes and a more complete use of fissile material. This capability significantly reduces nuclear waste and improves fuel sustainability compared to other Generation IV reactors. The metallic fuels, such as U-Zr and U-Pu-Zr, are attractive because of their higher thermal conductivity and higher density of fissile material, leading to improved breeding ratios and higher burnup rates. Through years of testing, SFRs, such as the EBR-II, could achieve a very high burnup (up to 750 GWD/tU) while traditional generation I to III+ reactors achieve around 30 GWD/tU. Since fast reactors, particularly SFRs, operate on a hard neutron spectrum, they utilize different geometries and cooling materials. This requires the use of different mechanistic models in nuclear codes to accurately capture the underlying physics. The NRC has recently shown interest in upgrading its diffusion codes to support the integration of SFRs. They have shown additional interest in improving the simulation capability of SFR metallic fuel, specifically U-10wt%Zr, even for high fuel burnups in excess of 10 at.%. </p><p dir="ltr">The purpose of this thesis is multifaceted. It serves to develop new mechanistic modelsto support the modeling and analysis of SFRs in existing nodal codes; it also serves to advance the current understanding of the principal effects of U-Zr fuel behavior during steady-state conditions. From the neutronics perspective, a new nodal method will be developed within the PARCS nodal code, along with generalized concepts such as reactivity feedback coefficients and thermal expansion, which will then be validated against the EBRII steady-state benchmark. From the fuel behavior perspective, mechanistic models that describe fuel redistribution, temperature distribution, fission gas, mechanical stresses, and point defect generation will be developed into the Purdue Fuel Performance (PFP) code for U-Zr fuel. The code will then be validated against the radial fuel concentration profile of two EBR-II U-Zr fuel pins.</p>