Thesis (M.S. in Engineering Science)--University of California, June 1963. / TID-4500 (19th Ed.). Bibliography: l. 70-71.

A safety analysis for the McMaster Nuclear Reactor has been carried out for postulated scenarios of loss or termination of forced flow in the reactor core in a state of shutdown, with loss of pool inventory of different magnitudes including core uncovery. Models were developed to evaluate the natural convection flow through the core assemblies for the different conditions within the aforementioned envelope. The flow rate was used to get the temperature or enthalpy rise along the heated channel in order to estimate the corresponding clad temperatures in the given scenarios. The models were constructed from first principles using the one-dimensional momentum conservation law, incorporating the Boussinesq approximation for the single-phase case and the Homogeneous Equilibrium Model assumptions when a two-phase mixture was present. In order to obtain the flow rate and enthalpy rise along the channel, knowledge of the assembly power and inlet temperature is required. The power was calculated using a well known decay power correlation. The pool temperature which was used as the assembly inlet temperature was calculated via a lumped parameter model using a simple energy balance between the core output (again by using the decay-heat profile) and the pool heatup. Heat losses from the pool were neglected and the model allowed for reaching saturation temperature in the pool. In this case, water vaporization was calculated using the latent heat to assess pool inventory loss rate. For all scenarios before core uncovery, the models predict that clad and fuel temperatures remained well below limits associated with clad blistering or melting. Consequently, it is asserted natural convection and acceptable temperatures will be sustained in the McMaster Nuclear Reactor while the core remains covered. In the most severe draining before uncovery, in which the pool drains to just before exposing the core, it takes approximately a week (180 hours) after shutdown for boiling to start in the core’s hottest channel. For core uncovery, the models predict that the clad remains below the blistering temperature for pool height at 9.4% of the heated channel’s height (corresponding to exposing about 61.7 cm of the assembly), and below melting temperature for pool height at 8.1% of the heated channel’s height (corresponding to exposing about 62.5 cm of the assembly). Both heights are below the height of the bottom of the lowest beam tube, at which the worst draining case will end. / Thesis / Master of Applied Science (MASc)

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 100-101). / The MIT Research Reactor (MITR-II) is currently undergoing analysis for the planned conversion from high enriched uranium (HEU) to low enriched uranium (LEU), as part of a global effort to minimize the availability of weapons-grade uranium. In support of efficient fuel management analysis with the new LEU fuel, a core design optimization tool has been developed. Using a coarse model, the tool can quickly consider the large range of refueling options available, and identify a solution which minimizes power peaking with the least fuel shuffling possible. The selected scheme can then be examined in greater detail with a more robust simulation tool. The unique geometry of the MITR core makes it difficult to develop a model that both runs very quickly and provides detailed power distribution information. Therefore, a correlation-based approach has been employed. Relationships between burnup, critical control blade position, core Um mass, and power distribution are used to predict fuel element U²³⁵ depletion, critical control blade motion, and power peaking. The tool applies the correlations to identify an optimal loading pattern, defined as the core which has the lowest maximum radial peaking factor in the set of valid solutions with the minimum number of fuel shuffling actions. The correlations that are utilized by the optimization tool were developed using data from simulations with MCODE-FM, a fuel management wrapper for the MCNP-ORIGEN linkage code MCODE. The correlations have been verified with results from additional MCODE-FM runs, and the code logic has been verified with the core loading solutions for a variety of input parameters. The verification found that the code is able to predict radial peaking, core mass, and general control blade motion with sufficient accuracy to develop a good refueling scheme. The tool provides the output solution in an interactive format, which allows the user to quickly examine small perturbations on the identified loading pattern. In addition to the optimization tool development, loading patterns for the mixed HEU-LEU fuel transition cores have been evaluated. This analysis identified general behavioral trends of the mixed-fuel cores, which serve as an initial basis for future transition core analysis. / by Heather M. Connaway. / S.M.

Em todo mundo, muitas instalações nucleares foram construídas e necessitarão serem desligadas em um determinado momento por estarem próximas do seu tempo recomendado de utilização que é de aproximadamente 40 anos. A AIEA (Agência Internacional de Energia Atômica) busca orientar e recomendar uma série de diretrizes para a realização de atividades de descomissionamento de instalações nucleares, com atenção especial aos países que não possuem um quadro regulatório legal que ampare as atividades de descomissionamento. O Brasil, até o momento, não possui uma norma específica que oriente as etapas de descomissionamento de reatores de pesquisa. Entretanto, em março de 2011 foi constituída uma comissão de estudo com a atribuição principal voltada às questões de descomissionamento das instalações nucleares brasileiras, culminando na resolução 133, de 8 de novembro de 2012, um projeto de norma que dispõe sobre o Descomissionamento de Usinas Nucleoelétricas. O Instituto de Pesquisas Energéticas e Nucleares (IPEN) possui dois reatores de pesquisa sendo um deles o reator IPEN/MB-01. O objetivo dessa dissertação de mestrado é elaborar um plano preliminar de descomissionamento desse reator de pesquisa, considerando a documentação técnica da instalação (RAS-Relatório de Análise de Segurança), as normas existentes da CNEN (Comissão Nacional de Energia Nuclear), assim como as recomendações da AIEA. Em termos de procedimentos de descomissionamento para reatores de pesquisa, este trabalho se baseou no que existe de mais moderno em experiências, estratégias e lições aprendidas realizadas e documentadas nas publicações da AIEA que abrangem técnicas e tecnologias de descomissionamento. Considerando estes conhecimentos técnicos e às peculiaridades da instalação, foi selecionada a estratégia de desmantelamento imediato, que corresponde ao inicio das atividades de descomissionamento assim que a instalação for desligada, dividindo-a em setores de trabalho. Como recurso de gerenciamento e acompanhamento do projeto de descomissionamento do reator e manutenção de registros, foi desenvolvido um banco de dados utilizando o programa Microsoft Access 2007, no qual contêm todos os itens e informações referentes ao plano preliminar de descomissionamento. O trabalho aqui descrito busca atender os requisitos, critérios técnicos e institucionais, incorporando o que se tem de mais atual em procedimentos de descomissionamento, podendo servir como guia para as demais instalações brasileiras. / Around the world, many nuclear plants were built and need to be turned off at a certain time because they are close to their recommended time of use is approximately 50 years. So the IAEA (International Atomic Energy Agency), seeks to guide and recommend a set of guidelines for the conduct of activities of nuclear facilities, with special attention to countries that do not have a framework regulatory Legal that sustain the activities of decommissioning. Brazil, so far, does not have a specific standard to guide the steps of the guidelines regarding decommissioning research reactors. However, in March 2011 a study committee was formed with the main task facing the issues of decommissioning of nuclear installations in Brazil, culminating in Resolution 133 of November 8, 2012, a standard project that treat about the Decommissioning of nucleoelectric plants. O Instituto de Pesquisas Energéticas e Nucleares (IPEN) has two research reactors one being the reactor IPEN/MB-01. The purpose of this master dissertation is to develop a preliminary plan for decommissioning this research reactor, considering the technical documentation of the facility (RAS-Safety Analysis Report), the existing standards of CNEN (National Nuclear Energy Commission), as well as IAEA recommendations. In terms of procedures for decommissioning research reactors, this work was based on what is most modern in experiences, strategies and lessons learned performed and documented in IAEA publications covering techniques and technologies for decommissioning. Considering these technical knowledges and due to the peculiarities of the facility, was selected to immediate dismantling strategy, which corresponds to the start of decommissioning activities once the installation is switched off, dividing it into work sectors. As a resource for monitoring and project management of reactor decommissioning and maintenance of records, we developed a database using Microsoft Access 2007, which contain all the items and information for the preliminary decommissioning plan. The work described here aims to meet the requirements, technical and institutional criteria, incorporating what is most current procedures and lessons learned of decommissioning, may serve as a guideline for the other brazilian facilities.

Em todo mundo, muitas instalações nucleares foram construídas e necessitarão serem desligadas em um determinado momento por estarem próximas do seu tempo recomendado de utilização que é de aproximadamente 40 anos. A AIEA (Agência Internacional de Energia Atômica) busca orientar e recomendar uma série de diretrizes para a realização de atividades de descomissionamento de instalações nucleares, com atenção especial aos países que não possuem um quadro regulatório legal que ampare as atividades de descomissionamento. O Brasil, até o momento, não possui uma norma específica que oriente as etapas de descomissionamento de reatores de pesquisa. Entretanto, em março de 2011 foi constituída uma comissão de estudo com a atribuição principal voltada às questões de descomissionamento das instalações nucleares brasileiras, culminando na resolução 133, de 8 de novembro de 2012, um projeto de norma que dispõe sobre o Descomissionamento de Usinas Nucleoelétricas. O Instituto de Pesquisas Energéticas e Nucleares (IPEN) possui dois reatores de pesquisa sendo um deles o reator IPEN/MB-01. O objetivo dessa dissertação de mestrado é elaborar um plano preliminar de descomissionamento desse reator de pesquisa, considerando a documentação técnica da instalação (RAS-Relatório de Análise de Segurança), as normas existentes da CNEN (Comissão Nacional de Energia Nuclear), assim como as recomendações da AIEA. Em termos de procedimentos de descomissionamento para reatores de pesquisa, este trabalho se baseou no que existe de mais moderno em experiências, estratégias e lições aprendidas realizadas e documentadas nas publicações da AIEA que abrangem técnicas e tecnologias de descomissionamento. Considerando estes conhecimentos técnicos e às peculiaridades da instalação, foi selecionada a estratégia de desmantelamento imediato, que corresponde ao inicio das atividades de descomissionamento assim que a instalação for desligada, dividindo-a em setores de trabalho. Como recurso de gerenciamento e acompanhamento do projeto de descomissionamento do reator e manutenção de registros, foi desenvolvido um banco de dados utilizando o programa Microsoft Access 2007, no qual contêm todos os itens e informações referentes ao plano preliminar de descomissionamento. O trabalho aqui descrito busca atender os requisitos, critérios técnicos e institucionais, incorporando o que se tem de mais atual em procedimentos de descomissionamento, podendo servir como guia para as demais instalações brasileiras. / Around the world, many nuclear plants were built and need to be turned off at a certain time because they are close to their recommended time of use is approximately 50 years. So the IAEA (International Atomic Energy Agency), seeks to guide and recommend a set of guidelines for the conduct of activities of nuclear facilities, with special attention to countries that do not have a framework regulatory Legal that sustain the activities of decommissioning. Brazil, so far, does not have a specific standard to guide the steps of the guidelines regarding decommissioning research reactors. However, in March 2011 a study committee was formed with the main task facing the issues of decommissioning of nuclear installations in Brazil, culminating in Resolution 133 of November 8, 2012, a standard project that treat about the Decommissioning of nucleoelectric plants. O Instituto de Pesquisas Energéticas e Nucleares (IPEN) has two research reactors one being the reactor IPEN/MB-01. The purpose of this master dissertation is to develop a preliminary plan for decommissioning this research reactor, considering the technical documentation of the facility (RAS-Safety Analysis Report), the existing standards of CNEN (National Nuclear Energy Commission), as well as IAEA recommendations. In terms of procedures for decommissioning research reactors, this work was based on what is most modern in experiences, strategies and lessons learned performed and documented in IAEA publications covering techniques and technologies for decommissioning. Considering these technical knowledges and due to the peculiarities of the facility, was selected to immediate dismantling strategy, which corresponds to the start of decommissioning activities once the installation is switched off, dividing it into work sectors. As a resource for monitoring and project management of reactor decommissioning and maintenance of records, we developed a database using Microsoft Access 2007, which contain all the items and information for the preliminary decommissioning plan. The work described here aims to meet the requirements, technical and institutional criteria, incorporating what is most current procedures and lessons learned of decommissioning, may serve as a guideline for the other brazilian facilities.

For the Texas A&M University Nuclear Science Center reactor a fuel depletion code was created to develop real-time fuel management capability. This code package links MCNP8 and ORIGEN26 and is interfaced through a Visual Basic code. Microsoft Visual Basic was used to create a user interface and for pre-and post-processing of MCNP and ORIGEN2 output. MCNP was used to determine the flux for all fuel and control rods within the core while ORIGEN2 used this flux along with the power history to calculate buildup and depletion for tracking the fuel isotopic evolution through time. A comparison of MCNP calculated fluxes and measured flux values were used to confirm the validity of the MCNP model. A comparison to Monteburns was used to add confidence to the correctness of the calculated fuel isotopics. All material isotopics were stored in a Microsoft Access database for integration with the Visual Basic code to allow for isotopics report generation for the Nuclear Science Center staff. This fuel management code performs its function with reasonable accuracy. It gathers minimal information from the user and burns the core over daily operation. After execution it stores all material data to the database for further use within NSCRFM or for isotopic report generation.

On Site Inspection is the final measure for verifying compliance of Member States with the Comprehensive Nuclear-Test-Ban Treaty. In order to enable the use of ³⁷Ar as a radiotracer for On Site Inspection, the sources of radioargon background must be characterized and quantified. A radiation transport model of the University of Texas at Austin Nuclear Engineering Teaching Laboratory (NETL) TRIGA reactor was developed to simulate the neutron flux in various regions of the reactor. An activation and depletion code was written to calculate production of ³⁷Ar in the facility based on the results of the radiation transport model. Results showed ³⁷Ar production rates of (6.567±0.31)×10² Bq·kWh⁻¹ in the re- actor pool and the air-filled irradiation facilities, and (5.811±0.40)×10⁴ Bq·kWh⁻¹ in the biological shield. Although ⁴⁰Ca activation in the biological shield was found to dominate the total radioargon inventory, the contribution to the effluent release rate would be diminished by the immobility of Ar generated in the concrete matrix and the long diffusion path of mobile radioargon. Diffusion of radioargon out of the reactor pool was found to limit the release rate but would not significantly affect the integrated release activity. The integrated ³⁷Ar release for an 8 hour operation at 950 kW was calculated to be (1.05±0.8)×10⁷ Bq, with pool emissions continuing for days and biological shield emissions continuing for tens of days following the operation. Sensitivity analyses showed that estimates for the time-dependent concentrations of ³⁷Ar in the NETL TRIGA could be made with the calculated buildup coefficients or through analytical solution of the activation equations for only (n,[gamma]) reactions in stable argon or (n,[alpha]) reactions in ⁴⁰Ca. Analyses also indicated that, for a generalized system, the integrated thermal flux can be used to calculate the buildup due to air activation and the integrated fast flux can be used to calculate the buildup due to calcium activation. Based on the results of the NETL TRIGA, an estimate of the global research reactor source term for ³⁷Ar and an estimate of ground-level ³⁷Ar concentrations near a facility were produced. / text

No description available.

L'objectif de cette thèse était d'aboutir à une meilleure compréhension du comportement thermique d'un calorimètre, différentiel non adiabatique permanent, dédié à la mesure de l'échauffement nucléaire en MTR, puis à une optimisation de ce capteur et des méthodes de mesures associées et enfin à une proposition d'une configuration calorimétrique « miniaturisée ». Du fait du principe même de la calorimétrie (quantification d'énergie à partir de mesures de températures), une approche analytique ciblant les aspects thermiques a été conduite. Cette thèse a consisté à la conception, au développement et à l'exploitation de nouveaux outils analytiques thermiques expérimentaux et numériques. Un modèle thermique 2D axisymétrique résolu par méthode des éléments finis (code CAST3M) a été mis en œuvre, validé et utilisé en conditions non irradiées ou irradiées dans le cadre d'une étude paramétrique complète portant sur la réponse de différentes configurations calorimétriques. Ces travaux ont permis le dimensionnement d'un capteur plus sensible adapté aux conditions ciblées lors des premières campagnes d'irradiation en périphérie du réacteur OSIRIS (< 2W/g). Ces travaux ont également permis de définir une nouvelle cellule calorimétrique à échange directionnel radial, plus compacte pour des expériences futures en cœur du réacteur RJH à fort échauffement nucléaire (20W/g). Un dispositif expérimental a été conçu afin d'étudier le calorimètre plus sensible pour différentes contraintes thermiques (puissance injectée, température du fluide caloporteur) et fluidiques (intensité de la convection forcée) imposées. / The aim of this thesis was to reach a better understanding of the calorimeter thermal behavior of a calorimeter, differential non-adiabatic permanent, dedicated to the nuclear heating measurement in MTR, an optimization of the sensor and measurement methods associated and finally a suggestion of a calorimetric configuration "miniaturized." Because of the principle of calorimetry (quantification of energy from temperature measurements), an analytical approach targeting thermal aspects was carried out. This thesis consisted of designing, development and exploitation of new experimental and numerical analytical thermal tools. A 2D axisymmetric thermal model solved by finite element method (CAST3M code) was implemented, validated and used in conditions unirradiated or irradiated through a complete parametric study related to the response of different calorimetric configurations. This work has allowed to design a more sensitive sensor adapted to conditions targeted for the first irradiation campaigns on the OSIRIS reactor reflector (<2W/g). These studies have also allowed to define a new calorimetric cell with radial directional exchange, more compact for future experiments in the RJH core with high nuclear heating (20W/g). An experimental set up was designed to study the most sensitive calorimeter in different thermal conditions (injected power, coolant temperature) and flow conditions (intensity of forced convection) imposed.

Sistemas pneumáticos de irradiação são instalações utilizadas em reatores nucleares de pesquisa. Sua função principal é de prover um meio rápido de envio de materiais para irradiação em posições localizadas nas proximidades do núcleo do reator. Durante sua utilização, cápsulas contendo os materiais de estudo são enviadas por meio de tubulações utilizando um fluido propulsor gasoso. Ao chegar à posição desejada, a cápsula sofre a exposição à radiação proveniente do reator possibilitando as transformações do material alocado em seu interior, porém como consequência da exposição também ocorre seu aquecimento térmico. Este trabalho estudou de forma numérica, utilizando a dinâmica dos fluidos computacional (CFD) e experimental, por meio de uma bancada de ensaios, o escoamento e transferência de calor durante o processo de irradiação. Os resultados encontrados demonstraram um aquecimento significativo para tempos de irradiação na ordem de 1 minuto considerando uma taxa de geração de calor constante, provocando a elevação da temperatura da cápsula a valores críticos para materiais de fabricação das cápsulas comumente utilizados como o polietileno de alta densidade (PEAD). Além disso, foram levantados os campos de velocidade, pressão e temperatura para o fluido propulsor e água de resfriamento no interior do tubo de irradiação que abriga a cápsula durante sua irradiação e avaliadas as respostas para diferentes modelos de turbulência nas simulações numéricas. Em função dos resultados obtidos concluiu-se que o estudo desenvolvido possibilitou exemplificar o processo de aquecimento das cápsulas e fornecer informações sobre as características do escoamento no interior do tubo de irradiação que abriga as cápsulas durante o processo de exposição. A utilização de diferentes modelos de turbulência nas simulações gerou resultados similares para o caso de estudo, porém pequenas variações em regiões de escoamento próximo à parede e em zonas de recirculação foram encontradas. / Pneumatic irradiation system facilities are used in nuclear research reactors. Its main function is to provide a fast means of sending materials to irradiation positions located near the reactor core. Capsules containing the sample materials are sent through pipes using a gaseous fluid propellant. Upon reaching the desired position, the capsule undergoes exposure to radiation from the reactor enabling the transformation of the material allocated inside, but as a consequence of exposure, its thermal heating also occurs. This study investigated numerically, using computational fluid dynamics (CFD), and experimentally the flow and heat transfer during the irradiation process. The results showed a significant heating for irradiation times on the order of 1 minute, considering a constant heat generation rate, thus causing increase in the capsule temperature up to critical values, for the materials that are commonly used for their manufacture. In particular, this is the case of the high density polyethylene (HDPE). Furthermore, the velocity, pressure and temperature fields were obtained for the propellant fluid and cooling water inside the irradiation tube house during its irradiation and the response of different turbulence modeling in the numerical simulations were analyzed. Based on the results obtained, it was possible to conclude that the developed study exemplified the heating process of the capsules and provided information about the characteristics of the flow inside the irradiation tube that houses the capsules during the exposure process. The use of different turbulence models in the simulations generated similar results for the study, however small variations in regions of flow near to the wall and inside recirculation zones were found.