Este trabalho tem como objetivo estudar as modificações introduzidas, ao longo de sucessivas versões, nos modelos empíricos do programa computacional FRAPCON utilizado para a simulação do comportamento sob irradiação de varetas combustíveis de Reatores a Água Leve Pressurizada (Pressurized Water Reactor - PWR) em regime de estado estacionário e sob condições de alta queima. No estudo, foram analisados os modelos empíricos utilizados pelo FRAPCON e que são apresentados em sua documentação oficial. Um estudo bibliográfico foi conduzido sobre os efeitos da alta queima em combustíveis nucleares visando melhorar o entendimento dos modelos utilizados pelo FRAPCON nestas condições. Foram feitas simulações do comportamento sob irradiação de uma vareta combustível típica de um reator PWR utilizando as versões 3.3, 3.4 e 3.5 do FRAPCON. Os resultados apresentados pelas diferentes versões do programa foram comparados entre si de forma a verificar as consequências das mudanças de modelo nos parâmetros de saída do programa. Foi possível observar que as modificações introduzidas trouxeram diferenças significativas nos resultados de parâmetros térmicos e mecânicos da vareta combustível, principalmente quando se evoluiu da versão FRAPCON-3.3 para a versão FRAPCON-3.5. Nessa ultima versão, obteve-se menores temperaturas na vareta combustível, menores tensões e deformações no revestimento, menor espessura da camada de oxido formada no revestimento a altas queimas na vareta combustível. / This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version.

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Made available in DSpace on 2014-10-09T12:33:40Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:06:26Z (GMT). No. of bitstreams: 0 / O Metano (CH4) é o segundo gás de efeito estufa mais importante, com aproximadamente 40% de sua emissão proveniente de fontes naturais, enquanto as fontes antrópicas representam cerca de 60%. Sua média global em 2009 foi de 1803ppb, que representa um aumento de 5ppb em relação ao ano anterior. Neste estudo foram calculados os fluxos de CH4, utilizando medidas de perfis verticais com aviões de pequeno porte, desde a superfície até 4,4km na Bacia Amazônica, sobre Santarém (SAN), Alta Floresta (ALF), Rio Branco (RBA) e Tabatinga (TAB), por meio do Método de Integração de Coluna. Estas medidas de CH4 em escala regional até o presente momento são únicas e representam uma nova abordagem nas emissões nesta escala. As medidas em SAN foram realizadas entre 2000 e 2010 e o fluxo de CH4 encontrado para este período foi de 53,0 ± 27,9mgCH4.m-2.dia-1. Para o ano de 2010, o maior fluxo de emissão de CH4 foi observado no lado leste da Bacia Amazônica, entre a costa e SAN, 56,4 ± 22,4mgCH4.m-2.dia-1. Entre a costa e ALF, ao sul da Bacia Amazônica, o fluxo médio anual foi de 17,1 ± 2,3mgCH4.m-2.dia-1, e entre a costa e os locais TAB e RBA, no lado oeste da Bacia, foi observado um fluxo médio anual de 18,7 ± 4,2 e 19,3 ± 10,2mgCH4.m-2.dia-1, respectivamente. Extrapolando os resultados obtidos em TAB e RBA para toda a área da Bacia Amazônica (5 milhões Km2) obtêm-se uma emissão de 34,7 ± 13,5TgCH4.ano-1. Com o objetivo de determinar a influência da queima de biomassa no fluxo regional de emissão de CH4, foi utilizada a correlação 6,4ppbCO/ppbCH4 calculada neste estudo, ALF foi o local de estudo que apresentou a maior influência no fluxo de CH4 oriundo da queima de biomassa, 23% do fluxo total anual. / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP

Submitted by Maria Eneide de Souza Araujo (mearaujo@ipen.br) on 2015-02-25T13:31:17Z No. of bitstreams: 0 / Made available in DSpace on 2015-02-25T13:31:17Z (GMT). No. of bitstreams: 0 / Fundação de Amparo à Pesquisa do Estado de São Paulo (FAPESP) / No panorama atual de mudanças climáticas, o Metano (CH4) é considerado o segundo principal gás de efeito estufa antrópico. Este trabalho teve como objetivo estudar o papel da Amazônia na emissão global de CH4, sendo esta a maior floresta tropical do mundo. Neste estudo foram realizados perfis verticais, utilizando aviões de pequeno porte, desde 150 m da superfície até 4,4 km, em quatro localidades da Bacia Amazônica, formando um grande quadrante abrangendo toda a Bacia. Os locais foram: próximo a Santarém (SAN; 2,8°S, 54,9°O), Alta Floresta (ALF; 8,8°S, 56,7°O), Rio Branco (RBA; 9,3°S, 67,6°O) e Tabatinga (TAB; 5,9°S, 70,0°O). Foram realizados quatro anos (2010-2013) de medidas contínuas em escala regional, quinzenalmente, totalizando 293 perfis verticais. Até o presente momento estas medidas são únicas e representam uma nova abordagem nas emissões nesta escala. Foram calculados os fluxos de CH4 nestas quatro localidades por meio do Método de Integração de Coluna e os fluxos anuais foram calculados através de média proporcional, considerando a área de influência de cada localidade. Os anos de 2010 e 2012 foram anos de seca, enquanto 2011 e 2013 foram anos com precipitação acima da média na Amazônia. Dos quatro anos de estudo apenas 2011 apresentou uma temperatura inferior a média. Os resultados obtidos mostraram que a Amazônia atua como uma importante fonte de CH4, com uma emissão de 25,4 Tg ano-1 (4% - 5% da emissão global), considerando a área da Amazônia Brasileira (4,2 milhões de km2). As emissões nesta região apresentaram variações regionais e anuais, com maiores emissões nos anos de seca. A emissão pela queima de biomassa não foi significativa nas regiões de estudo, enquanto as estimativas de emissões por fermentação entérica e manejo dos dejetos de animais foram significativas na maioria destas regiões. Os resultados obtidos ressaltam a importância da realização de estudos em escala regional para esclarecer o comportamento de toda a área da Bacia Amazônica Brasileira. / Tese (Doutorado em Tecnologia Nuclear) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP / FAPESP:11/17914-0

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Fuel burnup calculations provide material concentrations and intrinsic neutron and gamma source strengths as a function of irradiation and cooling time. Detailed, full-core 3D burnup calculations are critical for nuclear fuel management studies, including core design and spent fuel storage safety and safeguards analysis. For core design, specifically during refueling, full- core pin-wise, axially-dependent burnup distributions are necessary to determine assembly positioning to efficiently utilize fuel resources. In spent fuel storage criticality safety analysis, detailed burnup distributions enable best-estimate analysis which allows for more effective utilization of storage space. Additionally, detailed knowledge of neutron and gamma source distributions provide the ability to ensure nuclear material safeguards. The need for accurate and efficient burnup calculations has become more urgent for the simulation of advanced reactors and monitoring and safeguards of spent fuel pools. To this end, the Virginia Tech Transport Theory Group (VT3G) has been working on advanced computational tools for accurate modeling and simulation of nuclear systems in real-time. These tools are based on the Multi-stage Response-function Transport (MRT) methodology. For monitoring and safety evaluation of spent fuel pools and casks, the RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system has been developed. This dissertation presents a novel methodology and algorithm for performing 3D fuel bur- nup calculations, referred to as bRAPID- Burnup with RAPID . bRAPID utilizes the existing RAPID code system for accurate calculation of 3D fission source distributions as the trans- port calculation tool to drive the 3D burnup calculation. bRAPID is capable of accurately and efficiently calculating assembly-wise axially-dependent fission source and burnup dis- tributions, and irradiated-fuel properties including material compositions, neutron source, gamma source, spontaneous fission source, and activities. bRAPID performs 3D burnup calculations in a fraction of the time required by state-of-the-art methodologies because it utilizes a pre-calculated database of response functions. The bRAPID database pre-calculation procedure, and its automation, is presented. The ex- isting RAPID code is then benchmarked against the MCNP and Serpent Monte Carlo codes for a spent fuel pool and the U.S. Naval Academy Subcritical Reactor facility. RAPID is shown to accurately calculate eigenvalue, subcritical multiplication, and 3D fission source dis- tributions. Finally, bRAPID is compared to traditional, state-of-the art Serpent Monte Carlo burnup calculations and its performance will be evaluated. It is important to note that the automated pre-calculation proceedure is required for evaluating the performance of bRAPID. Additionally, benchmarking of the RAPID code is necessary to understand RAPID's ability to solve problems with variable burnups distributions and to asses its accuracy. / Ph. D. / In a nuclear reactor, the energy released from a fission reaction, the splitting of an atomic nucleus into smaller parts, is harnessed to generate electricity. Nuclear reactors rely on fuel, typically comprised of uranium oxide (UO₂). While the reactor is operating and the fuel is being used, or “burned”, for power production it is undergoing numerous nuclear reactions, including fission, and radioactive decays which alter the material composition. Knowing the time evolution of fuel as it is burned in the reactor, i.e., concentration of isotopes and sources of radiation, is critical. Nuclear reactor designers and operators use this information to optimize power production and perform safety analysis of used nuclear fuel. By performing fuel burnup calculations, material concentrations and radiation source strengths are obtained as a function of time in an operating nuclear reactor. Using traditional computational techniques, these calculations are extremely time consuming and, for certain problems, can be difficult to obtain an accurate solution. Ideally, a reactor designer would like to know the three-dimensional (3D) distribution of material compositions and sources; however this level of detail would require an excessive amount of calculation time, therefore simplified models and assumptions are used. For the design of the new generation of nuclear reactors, and monitoring and safeguards analysis, this level of detail will be required in lieu of the availability of experimental facilities which do not currently exist. This dissertation presents a novel methodology and algorithm for performing accurate 3D fuel burnup calculations in real-time, referred to as bRAPID (Burnup with RAPID). bRAPID utilizes an existing nuclear software, RAPID (Real-time Analysis for Particle transport and In-situ Detection), developed in the Virginia Tech Transport Theory Group (VT3G), which has been shown to accurately solve time-independent nuclear calculations in significantly less time than traditional approaches. bRAPID is capable of accurately calculating 3D material and source distributions as a function of time in an operating nuclear reactor, and requires significantly less time and computational resources than traditional approaches. To ensure that bRAPID is relatively easy to use, a number of automated routines have been developed and are presented. RAPID is benchmarked against the traditional code systems MCNP (Monte Carlo N-Particle) and Serpent, both of which are widely used in the nuclear community, for a spent fuel storage pool and the U.S. Naval Academy subcritical nuclear reactor facility. RAPID is shown to accurately calculate system parameters (eigenvalue and subcritical multiplication factor) and 3D fission source distributions. Finally, bRAPID is compared to the traditional burnup approach, using the Serpent code system. bRAPID is shown to accurately calculate system parameters and 3D material and source distributions in significantly less time than the traditional approach.

A fuel burnup model of a Pressurized Water Reactor (PWR) was developed based on one-group diffusion theory and used simple thermal cross sections. A computer program which simulates the depletion of the core of a PWR was written based on this model. The basic idea was to develop a fuel depletion program which could be readily understood by nuclear engineering students. Thus, accuracy was sacrificed for the sake of simplicity. The model was based upon a typical PWR with three concentric regions in the radial direction of differing fuel enrichment. Each of the regions was homogenized and the concentrations of the isotopes in each region were considered constant over a time interval. The isotopes considered were U-235, Pu-239, U-238, Xe-135, I-135, Sm-149, Pm-149 and the lumped burnable poison isotope. The flux was approximated by the sum of two trigonometric functions. The magnitude and shape of the flux were determined by holding power constant, constraining system to be critical and varying the soluble boron concentration to find the fla~test possible positive flux. A flux magnitude computed in this manner was compared to a similar flux magnitude given in a Final Safety Analysis Report. The concentrations of the isotopes were determined from the differential equations describing the rate of change of the concentrations. The behavior of the isotopes over core life was graphed and wherever possible compared to graphs from other sources. The concentrations calculated for U-235, U-238 and Pu-239 after 450 days were compared to the concentrations of the same isotopes calculated by a zero dimensional three-group model. The percentage difference between the concentrations determined by the two models varied from about 69% for Pu-239 to 1% for U-238. / Master of Science