Two new fuel assembly designs for light water reactors using advanced mixed-oxide fuels have been proposed to reduce the radiotoxicity of used nuclear fuel discharged from nuclear power plants. The research efforts of this thesis are the first to consider the effects of burnup on advanced mixed-oxide fuel assembly performance and thermal safety margin over an assembly?s expected operational burnup lifetime. In order to accomplish this, a new burnup-dependent thermal-hydraulic analysis methodology has been developed. The new methodology models many of the effects of burnup on an assembly design by including burnup-dependent variations in fuel pin relative power from neutronic calculations, assembly power reductions due to fissile content depletion and core reshuffling, and fuel material thermal-physical properties. Additionally, a text-based coupling method is developed to facilitate the exchange of information between the neutronic code DRAGON and thermal-hydraulic code VIPRE-01. The new methodology effectively covers the entire assembly burnup lifetime and evaluates the thermal-hydraulic performance against ANS Condition I, II, and III events with respect to the minimum departure from nucleate boiling ratio, peak cladding temperatures, and fuel centerline temperatures. A comprehensive literature survey on the thermal conductivity of posed fuel materials with burnup-dependence has been carried out to model the advanced materials in the thermal-hydraulic code VIPRE-01. Where documented conductivity values are not available, a simplified method for estimating the thermal conductivity has been developed. The new thermal conductivity models are based on established FRAPCON-3 fuel property models used in the nuclear industry, with small adjustments having been made to account for actinide additions. Steady-state and transient thermal-hydraulic analyses are performed with VIPRE- 01 for a reference UO2 assembly design, and two advanced mixed-oxide fuel assembly designs using the new burnup-dependent thermal-hydraulic analysis methodology. All three designs maintain a sufficiently large thermal margin with respect to the minimum departure from nucleate boiling ratio, and maximum cladding and fuel temperatures during partial and complete loss-of-flow accident scenarios. The presence of a thin (Am,Zr)O2 outer layer on the fuel pellet in the two advanced mixed-oxide fuel assembly designs increases maximum fuel temperatures during transient conditions, but does not otherwise greatly compromise the thermal margin of the new designs.

Gamma spectroscopy and gamma emission tomography are two non-destructive measurement techniques for assessing the performance of nuclear fuel which have been investigated in this thesis for existing and novel applications through theoretical studies and experimental demonstrations. For assessment of individual fuel rods using gamma spectroscopy, fuel assemblies are dismantled so that the fuel rods may be measured separately, which is time-consuming and may cause damage to the fuel. Gamma tomography is more seldom used, but its application on complete fuel assemblies would enable the assessment of individual fuel rods without the need to disassemble the fuel. Both techniques are based on recording gamma rays, emitted at characteristic energies from decaying radioactive products in the fuel. The feasibility of measuring short-lived fission gasses in the gas plenum of fuel rods with short cooling time was experimentally investigated. Based on the feasibility demonstration, a method was proposed and experimentally demonstrated for determining the fission gas release fraction of 133Xe in fuel rods with short cooling time. Additionally, a method for investigating the origin of released fission gasses based on the measured ratio of 133Xe/85Kr in the fuel rod gas plenum was demonstrated. These methods may be employed at research reactors, where fuel with short cooling time is available for measurement. A gamma emission tomography instrument has been designed, constructed and experimentally demonstrated on a Halden Reactor fuel assembly. Simulation studies showed that the instrument and the tomographic reconstruction methods employed may be useful for: identifying a leaking fuel rod in an assembly by its lack of fission gas content; reconstruction of the rod-wise fission product distributions in the fuel stack and plenum regions of the assembly; and determining the rod-wise fission gas release fractions. In the experimental demonstration, the rod-wise distributions of the fission products 137Cs and 85Kr in the fuel stack and plenum regions of the assembly were reconstructed, as well as the distributions of the activation products 60Co and 178mHf in the plenum region, revealing the plenum springs and tie rods, respectively. The reconstructed data was in the form of images, useful for qualitative assessment of the fuel.

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Cette thèse s'inscrit dans le cadre général de la tenue au séisme des coeurs de réacteurs nucléaires à eau pressurisée (REP). Plus précisément, l'objectif de cette thèse est l'étude expérimentale du couplage entre assemblages combustibles induit par un écoulement d'eau axial. Les phases de conception, réalisation et mise en service d'une nouvelle installation appelée ICARE EXPERIMENTAL sont présentées. ICARE EXPERIMENTAL a été conçue pour observer simultanément les vibrations de quatre maquettes d'assemblages combustibles (2x2) confinées sous écoulement ascendant. Une nouvelle méthode d'analyse de données combinant analyse temps-fréquence et décomposition sur modes propres orthogonaux (POD) est décrite. Cette méthode, appelée Sliding Window POD (SWPOD), permet l'analyse de signaux à plusieurs composantes dont la répartition spatiale de l'énergie et le contenu fréquentiel varient avec le temps. Dans le cas de systèmes mécaniques (linéaires et non-linéaires), le lien entre les modes propres orthogonaux obtenus par la SWPOD et les modes normaux (linéaires et non-linéaires) est étudié. Les mesures obtenues avec l'installation ICARE EXPERIMENTAL sont analysées avec la SWPOD. Les premiers résultats mettent en évidence des mouvements caractéristiques des assemblages non excités, au passage de leurs résonances. Ce couplage entre assemblages combustibles, induit par le fluide, est reproduit par les simulations réalisées à l'aide du code de calcul COEUR3D. Ce code est basé sur une approche milieu poreux pour simuler un réseau d'assemblages combustibles sous écoulement. / The present study is in the scope of pressurized water reactors (PWR) core response to earthquakes. The goal of this thesis is to measure the coupling between fuel assemblies caused an axial water flow. The design, production and installation a new test facility named ICARE EXPERIMENTAL are presented. ICARE EXPERIMENTAL was built in order to measure simultaneously the vibrations of four fuel assemblies (2x2) under an axial flow. A new data analysis method combining time-frequency analysis and orthogonal mode decomposition (POD) is described. This method, named Sliding Window POD (SWPOD), allows analysing multicomponent data, of which spatial repartition of energy and frequency content are time dependent. In the case of mechanical systems (linear and nonlinear), the link between the proper orthogonal modes obtained through SWPOD and the normal modes (linear and nonlinear) is studied. The measures acquired with the ICARE EXPERIMENTAL installation are analysed using the SWPOD. The first results show characteristic behavior of the free fuel assemblies at their resonances. The coupling between fuel assemblies, induced by the fluid, is reproduced by simulations performed using the COEUR3D code. This code is based on a porous media model in order to simulate a fuel assemblies network under axial flow.

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PWR, fuel assembly, benchmark, burnup, lattice, SCALE, Polaris, validation, reactivity

Submitted by Marco Antonio Oliveira da Silva (maosilva@ipen.br) on 2016-12-21T17:48:50Z No. of bitstreams: 0 / Made available in DSpace on 2016-12-21T17:48:50Z (GMT). No. of bitstreams: 0 / Após o aumento de potência do reator IEA-R1 de 2 MW para 5 MW observou-se um aumento da taxa de corrosão nas placas laterais de alguns elementos combustíveis e algumas dúvidas surgiram com relação ao valor de vazão utilizada nas análises termo-hidráulicas. A fim de esclarecer e medir a distribuição de vazão real pelos elementos combustíveis que compõe o núcleo do reator IEA-R1, um elemento combustível protótipo, sem material nuclear, chamado DMPV-01 (Dispositivo para Medida de Pressão e Vazão), em escala real, foi projetado e construído em alumínio. A vazão no canal entre dois elementos combustíveis é muito difícil de estimar ou ser medida. Esta vazão é muito importante no processo de resfriamento das placas laterais. Este trabalho apresenta a concepção e construção de um elemento combustível instrumentado para medir a temperatura real nestas placas laterais para melhor avaliar as condições de resfriamento do combustível. Quatorze termopares foram instalados neste elemento combustível instrumentado. Quatro termopares em cada canal lateral e quatro no canal central, além de um termopar no bocal de entrada e outro no bocal de saída do elemento. Existem três termopares para medida de temperatura do revestimento e um para a temperatura do fluido em cada canal. Três séries de experimentos, para três configurações distintas, foram realizadas com o elemento combustível instrumentado. Em dois experimentos uma caixa de alumínio foi instalada ao redor do núcleo para reduzir o escoamento transverso entre os elementos combustíveis e medir o impacto na temperatura das placas externas. Dada a tamanha quantidade de informações obtidas e sua utilidade no projeto, melhoria e capacitação na construção, montagem e fabricação de elementos combustíveis instrumentados, este projeto constitui um importante marco no estudo de núcleos de reatores de pesquisa. As soluções propostas podem ser amplamente utilizadas para outros reatores de pesquisa. / Tese (Doutorado em Tecnologia Nuclear) / IPEN/T / Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP