O reator de pesquisas IEA-R1 opera em regime de 40 h semanais à potência de 4,5 MW. Nestas condições, os cestos disponíveis para o armazenamento dos elementos combustíveis irradiados possuem menos de metade da sua capacidade inicial. Assim, nestas condições de operação, teremos apenas cerca de seis anos de capacidade para armazenamento. Considerando que a vida útil desejada do IEA-R1 seja de pelo menos mais 20 anos, será necessário aumentar a capacidade de armazenamento de combustível irradiado. Dr. Henrik Grahn, especialista da Agência Internacional de Energia Atômica sobre o armazenamento molhado (em piscinas de estocagem), ao visitar o reator IEA-R1 (setembro/2012) fez algumas recomendações. Entre elas, a concepção e instalação de cestos fabricados com aço inoxidável borado e internamente revestidos com uma película de alumínio, de modo que a corrosão dos elementos combustíveis não ocorresse. Após uma revisão da literatura sobre opções de materiais disponíveis para esse tipo de aplicação chegamos ao BoralcanTM fabricado pela 3M devido suas propriedades. Este trabalho apresenta estudos sobre a análise de criticalidade com o código computacional MCNP-5 utilizando duas bibliotecas americanas de dados nucleares avaliados (ENDF/B-VI e ENDF/B-VII) comparativamente. Estas análises demonstraram a possibilidade de dobrar a capacidade de armazenamento de elementos combustíveis, no mesmo espaço ocupado pelos cestos atuais, atendendo a demanda do reator de pesquisas IEA-R1 e também satisfazendo os requisitos de segurança da Comissão Nacional de Energia Nuclear (CNEN) e da Agência Internacional de Energia Atômica (IAEA). / The IEA-R1 research reactor on business under the 40 h weekly to the power of 4.5 MW. Under these conditions, the racks available for the storage of spent fuel elements have less than half of its initial capacity. Thus, in these operating conditions, we will have only about six years of capacity for storage. Whereas the desired service life of the IEA-R1 is at least another 20 years it will be necessary to increase the storage capacity of spent fuel. Dr. Henrik Grahn, an expert of the International Atomic Energy Agency on the wet storage (in storage pools), to visit the IEA-R1 reactor (September/2012) made some recommendations. Among them, the design and installation of baskets made of borated stainless steel and internally lined with an aluminum film, so that corrosion of the fuel elements does not occur. After a literature review of material options available for this type of use, we got to BoralcanTM manufactured by 3M due to its properties. This work presents studies on the criticality analysis with the computer code MCNP-5 using two American libraries of the Evaluated Nuclear Data (ENDF/B-VI and ENDF/BVII), and compare results based on each database. These analyzes demonstrated the possibility of doubling the storage capacity of fuel elements in the same space occupied by the current baskets, meeting the demand of the IEA-R1 research reactor and also meeting the security requirements and of the National Commission of Nuclear Energy (CNEN) and of the International Atomic Energy Agency (IAEA).

O reator de pesquisas IEA-R1 opera em regime de 40 h semanais à potência de 4,5 MW. Nestas condições, os cestos disponíveis para o armazenamento dos elementos combustíveis irradiados possuem menos de metade da sua capacidade inicial. Assim, nestas condições de operação, teremos apenas cerca de seis anos de capacidade para armazenamento. Considerando que a vida útil desejada do IEA-R1 seja de pelo menos mais 20 anos, será necessário aumentar a capacidade de armazenamento de combustível irradiado. Dr. Henrik Grahn, especialista da Agência Internacional de Energia Atômica sobre o armazenamento molhado (em piscinas de estocagem), ao visitar o reator IEA-R1 (setembro/2012) fez algumas recomendações. Entre elas, a concepção e instalação de cestos fabricados com aço inoxidável borado e internamente revestidos com uma película de alumínio, de modo que a corrosão dos elementos combustíveis não ocorresse. Após uma revisão da literatura sobre opções de materiais disponíveis para esse tipo de aplicação chegamos ao BoralcanTM fabricado pela 3M devido suas propriedades. Este trabalho apresenta estudos sobre a análise de criticalidade com o código computacional MCNP-5 utilizando duas bibliotecas americanas de dados nucleares avaliados (ENDF/B-VI e ENDF/B-VII) comparativamente. Estas análises demonstraram a possibilidade de dobrar a capacidade de armazenamento de elementos combustíveis, no mesmo espaço ocupado pelos cestos atuais, atendendo a demanda do reator de pesquisas IEA-R1 e também satisfazendo os requisitos de segurança da Comissão Nacional de Energia Nuclear (CNEN) e da Agência Internacional de Energia Atômica (IAEA). / The IEA-R1 research reactor on business under the 40 h weekly to the power of 4.5 MW. Under these conditions, the racks available for the storage of spent fuel elements have less than half of its initial capacity. Thus, in these operating conditions, we will have only about six years of capacity for storage. Whereas the desired service life of the IEA-R1 is at least another 20 years it will be necessary to increase the storage capacity of spent fuel. Dr. Henrik Grahn, an expert of the International Atomic Energy Agency on the wet storage (in storage pools), to visit the IEA-R1 reactor (September/2012) made some recommendations. Among them, the design and installation of baskets made of borated stainless steel and internally lined with an aluminum film, so that corrosion of the fuel elements does not occur. After a literature review of material options available for this type of use, we got to BoralcanTM manufactured by 3M due to its properties. This work presents studies on the criticality analysis with the computer code MCNP-5 using two American libraries of the Evaluated Nuclear Data (ENDF/B-VI and ENDF/BVII), and compare results based on each database. These analyzes demonstrated the possibility of doubling the storage capacity of fuel elements in the same space occupied by the current baskets, meeting the demand of the IEA-R1 research reactor and also meeting the security requirements and of the National Commission of Nuclear Energy (CNEN) and of the International Atomic Energy Agency (IAEA).

This work aims to develop a theoretical power operations envelope for the PUR-1 reactor. Given the bulk coolant temperature, the reactor’s power level is limited primarily by the Onset of Nucleate Boiling. Additional limitations to the reactor power are explored including the dose rate at the top of the pool due to shine and the airborne effluent of argon and nitrogen. Operations in excess of the facility cooling capacity will be proposed and are already permitted at other US research reactor facilities, provided temperature limitations are met. The MCNP and NATCON code packages have been implemented to assist in power limitation measurement. A brief discussion on the licensing considerations is included to provide some framework for pursuit of these higher power levels. The maximum power consideration ensures continued full use of the facility while maximizing its effectiveness in the teaching laboratories and access to researchers. The final power level is limited by the administrative dose limit at the top of the reactor pool as well as the Onset of Nucleate Boiling power level as a function of bulk pool temperature. The result is an operational envelope which would allow operators to have the maximum neutron flux without changing the facility or creating phase transition within the light water coolant.

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.

The National Bureau of Standards Reactor (NBSR) located on the National Institute of Standards and Technology (NIST) Gaithersburg campus, is currently underway of fuel conversion from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. One particular challenging part of the conversion of the NBSR is the high average flux level (2.5×1014 n/cm2-s) required to maintain experimental testing capabilities of the reactor, without significant changes to the external structures of the reactor. Recently the General Atomics (GA) Training Research Isotopes General Atomics (TRIGA) fuel has shown some promising features as a LEU candidate for the high performance research reactors such as the NBSR. The GA fuel has a long history of success in conversion of research reactors since it was developed in 1980s. The UZrH compound in the GA fuel has seen success in long term TRIGA reactors, and is a proven safe LEU alternative. This study performs a neutronics evaluation of the TRIGA fuel under the schema of the NBSR’s heavy conversion requirements in order to examine whether the TRIGA fuel is a viable option for conversion of the NBSR. To determine the most optimal path of conversion, we performed a feasibility study with particular regard to the fuel dimensions, fuel rod configurations, cladding, as well as fuel structure selection. Based on the outcome of the feasibility study, an equilibrium core is then generated following the NBSR’s current fuel management schema. Key neutronics performance characteristics including flux distribution, power distribution, control rod (i.e., shim arms) worth, as well as kinetics parameters of the equilibrium core are calculated and evaluated. MCNP6, a Monte Carlo based computational modeling software was intensively used to aid in these calculations. The results of this study will provide important insight on the effectiveness of conversion, as well as determine the viability of the conversion from HEU to LEU using the GA fuel.

Thesis (Ph. D.)--University of Missouri-Columbia, 2002. / Typescript. Vita. Includes bibliographical references. Also available on the Internet.

Thesis (Ph. D.)--University of Missouri-Columbia, 2002. / Typescript. Vita. Includes bibliographical references. Also available on the Internet.

No description available.

Pre-detonation technical nuclear forensics techniques for research reactor spent fuel were developed in a collaborative project with Savannah River National Lab ratory. An inverse analysis method was employed to reconstruct reactor parameters from a spent fuel sample using results from a radiochemical analysis. In the inverse analysis, a reactor physics code is used as a forward model. Verification and validation of different reactor physics codes was performed for usage in the inverse analysis. The verification and validation process consisted of two parts. The first is a variance analysis of Monte Carlo reactor physics burnup simulation results. The codes used in this work are MONTEBURNS and MCNPX/CINDER. Both utilize Monte Carlo transport calculations for reaction rate and flux results. Neither code has a variance analysis that will propagate through depletion steps, so a method to quantify and understand the variance propagation through these depletion calculations was developed. The second verification and validation process consisted of comparing reactor physics code output isotopic compositions to radiochemical analysis results. A sample from an Oak Ridge Research Reactor spent fuel assembly was acquired through a drilling process. This sample was then dissolved in nitric acid and diluted in three different quantities, creating three separate samples. A radiochemical analysis was completed and the results were compared to simulation outputs at different levels ofdetail. After establishing a forward model, an inverse analysis was developed to re-construct the burnup, initial uranium isotopic compositions, and cooling time of a research reactor spent fuel sample. A convergence acceleration technique was used that consisted of an analytical calculation to predict burnup, initial 235U, and 236U enrichments. The analytic calculation results may also be used stand alone or in a database search algorithm. In this work, a reactor physics code is used as a for- ward model with the analytic results as initial conditions in a numerical optimization algorithm. In the numerical analysis, the burnup and initial uranium isotopic com- positions are reconstructed until the iterative spent fuel characteristics converge with the measured data. Upon convergence of the sample’s burnup and initial uranium isotopic composition, the cooling time can be reconstructed. To reconstruct cooling time, the standard decay equation is inverted and solved for time. Two methods were developed. One method uses the converged burnup and initial uranium isotopic compositions along in a reactor depletion simulation. The second method uses an isotopic signature that does not decay out of its mass bin and has a simple production chain. An example would be 137Cs which decays into the stable 137Ba. Similar results are achieved with both methods, but extended shutdown time or time away from power results in over prediction of the cooling time. The over prediction of cooling time and comparison of different burnup reconstruction isotope results are indicator signatures of extended shutdown or time away from power. Due to dynamic operation in time and function, detailed power history reconstruction for research reactors is very challenging. Frequent variations in power, repeated variable shutdown time length, and experimentation history affect the spectrum an individual assembly is burned with such that full reactor parameter reconstruction is difficult. The results from this technical nuclear forensic analysis may be used with law enforcement, intelligence data, macroscopic and microscopic sample characteristics in a process called attribution to suggest or exclude possible sources of origin for a sample.

Os elementos de combustível de um reator nuclear de pesquisa tipo MTR (\"Material Testing Reactor\") são, em sua grande maioria, formados por placas de combustível revestidas com alumínio contendo no cerne silicileto de urânio (U3Si2) disperso em matriz de alumínio. Essas placas possuem espessura da ordem de milímetros e comprimentos muito maiores em relação à sua espessura. Elas são dispostas paralelamente no conjunto que forma o elemento combustível, de maneira a formar canais entre elas com poucos milímetros de espessura, por onde escoa o fluido de refrigeração (água leve ou água pesada). Essa configuração, associada à necessidade de um escoamento com altas vazões para garantir o resfriamento das placas em operação, pode gerar problemas de falhas mecânicas das placas de combustível devido às vibrações induzidas pelo escoamento nos canais e, consequentemente, acidentes de proporções graves no caso de velocidade crítica que possa gerar o colapso das placas. Embora não haja ruptura das placas de combustível durante o colapso, as deflexões permanentes excessivas das placas podem causar bloqueio do canal de escoamento no núcleo do reator e levar ao superaquecimento nas placas. Para este trabalho, foram desenvolvidas uma bancada experimental com capacidade para altas vazões volumétricas (Q=100 m3/h) e uma seção de testes que simula um elemento combustível do tipo placa com três canais de resfriamento. A seção de testes foi construída com placas de alumínio e acrílico e foi instrumentada com sensores de deformação, sensores de pressão, um acelerômetro e um tubo de pitot. As dimensões da seção de testes foram baseadas nas dimensões do Elemento Combustível do Reator Multipropósito Brasileiro (RMB), cujo projeto está sendo coordenado pela Comissão Nacional de Energia Nuclear - CNEN. Os experimentos realizados alcançaram o objetivo de chegar à condição de velocidade crítica de Miller com o colapso das placas. A velocidade crítica foi atingida com 14,5 m/s levando a consequente deformação plástica das placas que formam o canal do escoamento. O canal central na entrada da seção de testes apresentou uma abertura de 3 mm em seu centro, causando um grande bloqueio do escoamento nos canais laterais. Este comportamento foi v constatado visualmente durante a desmontagem da seção de testes, ilustrado e discutido na análise de resultados apresentado neste trabalho. O bloqueio dos canais também foi observado por meio de gráficos de queda de pressão e por gráficos das deformações da entrada, centro e saída das placas contra a velocidade média da seção de testes. Observou-se uma queda da resistência hidráulica da seção de testes devido ao aumento da seção transversal de escoamento no canal central e um aumento exponencial das deformações quando da ocorrência da velocidade crítica. Comparativamente, o valor experimental obtido para velocidade crítica na seção de testes foi da ordem de 85% do valor obtido por cálculo com a expressão teórica de Miller. Os experimentos realizados permitiram um melhor entendimento da interação fluido estrutura em elementos de combustível tipo placa como: valores de frequências de vibrações naturais, instabilidade fluido elástica e desenvolvimento de técnicas para a detecção de valores de velocidade crítica. / The fuel elements of a MTR (Material Testing Reactor) type nuclear reactor are mostly composed of aluminum-coated fuel plates containing the core of uranium silica (U3Si2) dispersed in an aluminum matrix. These plates have a thickness of the order of millimeters and are much longer in relation to their thickness. They are arranged in parallel in the assembly forming the fuel element to form channels between them a few millimeters in thickness, through which there is a flow of the coolant (light water or heavy water). This configuration, combined with the need for a flow at high flow rates to ensure the cooling of the fuel element in operation, may create problems of mechanical failure of fuel plate due to the vibration induced by the flow in the channels. In the case of critical velocity may cause collapse of the plates. Although there is no rupture of the fuel plates during collapse, excessive permanent deflections of the plates can cause blockage of the flow channel in the reactor core and lead to overheating in the plates. For this study were developed an experimental bench capable of high volume flows (Q = 100 m3/h) and a test section that simulates a plate-like fuel element with three cooling channels. The test section was constructed with aluminum and acrylic plates and was instrumented with straingauge sensors, pressure sensors, accelerometer and a tube of pitot. The dimensions of the test section were based on the dimensions of the Fuel Element of the Brazilian Multipurpose Reactor (RMB), whose project is being coordinated by the National Commission of Nuclear Energy (CNEN). The experiments performed attained the objective of reaching Miller\'s critical velocity condition with the collapse of the plates. The critical velocity was reached with 14.5 m/s leading to the consequent plastic deformation of the plates forming the flow channel. The central channel had a 3mm aperture in its center, causing a large blockage of the flow in the lateral channels. This behavior was observed visually during the disassembly of the test section, illustrated and discussed in the results analysis presented in this work. Blocking of the channels was also observed by means of graphs of pressure drop and graphs of the deformations of the entrance, center and exit of the plates against the average speed vii of the section of tests. It was observed a decrease of the hydraulic resistance of the section of tests due to the increase of the transversal section of flow in the central channel and an exponential increase of the deformations when the critical speed occurrence. Comparatively, the value obtained for critical velocity in the test section through the experiments was of the order of 85% of the value obtained by calculation with Miller\'s theoretical expression. The experiments allowed a better understanding of the structure fluid interaction in plate type fuel elements such as: natural vibration frequency values, elastic fluid instability and development of techniques for the detection of critical velocity values.