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Master of Science / Department of Mechanical and Nuclear Engineering / Jeremy Roberts / The reactivity method is an indirect nondestructive technique to estimate integral burnup in fuel elements. In this method, the assumption is made that reactivity worth of a fuel element is a known function of burnup, often a linear relationship. When a fuel element burns, reactivity is reduced due to depletion of fissile actinides and generation of neutron-absorbing fission products. Currently, there is a lack of experimental data to verify the current composition of the KSU TRIGA (Training Research Isotopes General Atomics) fuel. Moreover, the KSU TRIGA Mark II staff method of estimating burnup is admittedly inaccurate due to its simple approximations. This work presents the positive period technique as convenient method use only the excess reactivity of the KSU core to compute reactivity via the inhour equation. Period measurements are determined via extraction and manipulation of the time dependent power data in the measurements. MCNP and Serpent modeling codes are both used extract the neutron kinetics parameters necessary in the inhour equation. Seven axial discretization of the KSU fuel was modeled, which minimizes the reactivity biases as function of burnup. Moreover, two unit cell models of the KSU TRIGA fuel were investigated. Modeled reactivity worths were computed using the KCODE in MCNP for comparative analysis. The burnup steps using two power peaking factor methods were developed to account for the biases introduced initial burnup of fuel prior to installation at KSU. By using the error distribution given by the two method to generate 200 test cases of the burnup steps can yield to reactivity worths as a function of burnup with quantifiable uncertainties. Finally, the results suggest that validation from another nondestructive technique such as gamma spectroscopy is necessary to asses the reactivity biases observed for higher burnup fuel elements due to unknown radial orientations. This work ultimately supports the production of a high-fidelity model of the KSU reactor.

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.

If international agreements regarding the need to significantly reduce greenhouse gas emissions are to be met then there is a high probability that the shipping industry will have to reduce its greenhouse gas emissions. For emission reductions from ships greater than around 40% then alternatives to fossil fuels - such as nuclear energy - will very likely be required. Whilst nuclear powered ships have successfully operated at sea for a number of decades, these have been primarily naval systems (or derivatives of naval systems such as icebreakers) and a few demonstration projects using reactors with low power outputs. The operational requirement for large civilian vessels (for example high capacity factors and limited personnel) mean the naval and past demonstration reactor systems are ill-suited for use in the current fleet of commercial container ships. There have been relatively few studies performed addressing the likely requirements upon core design a marine reactor would have to meet. This study addresses those issues and also implements a Pressurised Water Reactor core design capable of achieving these requirements. Furthermore, in order to simplify reactor operation for a limited number of personnel on board, the chemical reactivity control system has been eliminated during power operation. This has resulted in a novel low power density core that does not require refuelling for 15 years. The neutronic and fuel performance behaviour of this system has been studied with conventional UO2 fuel and thorium-uranium oxide ((Th,U)O2) fuel. With respect to (Th,U)O2 fuel there has been limited analysis comparing the performance of key fuel characteristics, such as fission gas release and thermal conductivity, as a function of uranium content in (Th,U)O2 fuel and their impact on fuel behaviour. Furthermore, the performance of neutronic codes for modelling Th-232 and U-233 from a variety of experiments using modern nuclear data libraries (post 1990) is lacking. Both of these issues are addressed in this study. Whilst it is frequently stated that thorium-based oxide fuel is superior to UO2 fuel it was found that due to the sensitivity of thermal conductivity on temperature and uranium content this was not true for the core designed in this study. The (Th,U)O2 core showed no net economic benefits with respect to the UO2 core and it was found that the fuel performance of (Th,U)O2 fuel was worse than the UO2 fuel in the reactor designed here. The UO2 core design, however, was able to satisfactorily meet the majority of requirements placed upon the system.

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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) / Tese (Doutorado em Tecnologia Nuclear) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP / FAPESP:11/17914-0