<p>Presently there is a clear trend of increasing demands on in-pile performance of nuclear fuel. Higher target burnups, part length rods and various fuel additives are some examples of this trend. Together with an increasing demand from the public for even safer nuclear power utilisation, this implies an increased focus on various experimental, preferably non-destructive, methods to characterise the fuel.</p><p>This thesis focuses on the development and experimental evaluation of such methods. In its first part, the thesis presents a method based on gamma-ray spectroscopy with germanium detectors that have been used at various power reactors in Europe. The aim with these measurements is to provide information about the thermal power distribution within fuel assemblies in order to validate core physics production codes. The early closure of the Barsebäck 1 BWR offered a unique opportunity to perform such validations before complete depletion of burnable absorbers in Gd-rods had taken place. To facilitate the measurements, a completely submersible measuring system, LOKET, was developed allowing for convenient in-pool measurements to be performed.</p><p>In its second part, the thesis describes methods that utilise in-pile measurements. These methods have been used in the Halden test-reactor for determination of fission gas release, pellet-cladding interaction studies and fuel development studies.</p><p>Apart from the power measurements, the LOKET device has been used for fission gas release (FGR) measurements on single fuel rods. The significant reduction in fission gas release in the modern fuel designs, in comparison with older designs, has been demonstrated in a series of experiments. A FGR database covering a wide range of burnup, power histories and fuel designs has been compiled and used for fuel performance analysis. The fission gas release has been measured on fuel rods with average burnups well above 60 MWd/kgU. The comparison between core physics calculations (PHOENIX-4/POLCA-7) and the in-pool measurements of thermal power indicates that the nodal power can generally be predicted with an accuracy within 4% and the bundle power with an accuracy better than 2%, expressed as rms errors.</p><p>In-pile experiments have successfully simulated the conditions that occur in a fuel rod following a primary debris failure, being secondary fuel degradation. It was concluded that massive hydrogen pick-up takes place during the first few days following the primary failure and that a pre-oxidized layer does not function as a barrier towards hydriding in an environment with a very high partial pressure of hydrogen. Another series of in-pile experiments clearly indicate that increased UO₂ grain size is an effective way of suppressing fission gas release in LWR fuel up to the burnup level covered (55 MWd/kgUO₂).</p>

Presently there is a clear trend of increasing demands on in-pile performance of nuclear fuel. Higher target burnups, part length rods and various fuel additives are some examples of this trend. Together with an increasing demand from the public for even safer nuclear power utilisation, this implies an increased focus on various experimental, preferably non-destructive, methods to characterise the fuel. This thesis focuses on the development and experimental evaluation of such methods. In its first part, the thesis presents a method based on gamma-ray spectroscopy with germanium detectors that have been used at various power reactors in Europe. The aim with these measurements is to provide information about the thermal power distribution within fuel assemblies in order to validate core physics production codes. The early closure of the Barsebäck 1 BWR offered a unique opportunity to perform such validations before complete depletion of burnable absorbers in Gd-rods had taken place. To facilitate the measurements, a completely submersible measuring system, LOKET, was developed allowing for convenient in-pool measurements to be performed. In its second part, the thesis describes methods that utilise in-pile measurements. These methods have been used in the Halden test-reactor for determination of fission gas release, pellet-cladding interaction studies and fuel development studies. Apart from the power measurements, the LOKET device has been used for fission gas release (FGR) measurements on single fuel rods. The significant reduction in fission gas release in the modern fuel designs, in comparison with older designs, has been demonstrated in a series of experiments. A FGR database covering a wide range of burnup, power histories and fuel designs has been compiled and used for fuel performance analysis. The fission gas release has been measured on fuel rods with average burnups well above 60 MWd/kgU. The comparison between core physics calculations (PHOENIX-4/POLCA-7) and the in-pool measurements of thermal power indicates that the nodal power can generally be predicted with an accuracy within 4% and the bundle power with an accuracy better than 2%, expressed as rms errors. In-pile experiments have successfully simulated the conditions that occur in a fuel rod following a primary debris failure, being secondary fuel degradation. It was concluded that massive hydrogen pick-up takes place during the first few days following the primary failure and that a pre-oxidized layer does not function as a barrier towards hydriding in an environment with a very high partial pressure of hydrogen. Another series of in-pile experiments clearly indicate that increased UO2 grain size is an effective way of suppressing fission gas release in LWR fuel up to the burnup level covered (55 MWd/kgUO2).

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Made available in DSpace on 2014-10-09T12:34:00Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:01:07Z (GMT). No. of bitstreams: 0 / O trabalho teve como objetivo a investigação de Metodologias de Cálculo dos Reatores Subcríticos acionados por fonte externa de nêutrons, tais como, \"Accelerator Driven Subcritical Reactor\" (ADSR) e \"Fusion Driven Subcritical Reator\" (FDSR) , que são reatores nucleares subcríticos com uma fonte externa de nêutrons. Tais nêutrons são produzidos, no caso do ADSR, através da interação de partículas aceleradas (prótons, deutério) com um alvo (Pb, Bi, etc) ou através das reações de fusão, no caso do FDSR. Este conceito de reator vem sendo objeto de intensa pesquisa, sobretudo pela possibilidade de ser utilizado para transmutar o enorme inventario de rejeitos nucleares, principalmente os transurânicos (TRU) e os produtos de fissão de meia-vida longa (LLFP). Neste trabalho enfatiza os seguintes aspectos: (i) complementar e aprimorar a metodologia de cálculos neutrônicos com queima e transmutação e implementá-la computacionalmente; (ii) e utilizando esta metodologia, participar dos Projetos Coordenados de Pesquisa (CRP) da Agência Internacional de energia Atômica \"Analytical and Experimental Benchmark Analysis of ADS\" e \"Collaborative work on use of LEU in ADS\", principalmente na reprodução dos resultados experimentais da instalação subcrítica Yalina Booster e também no cálculo de um núcleo subcrítico do reator IPEN/MB-01, (iii) analisar comparativamente diferentes bibliotecas de dados nucleares, no cálculo de parâmetros integrais (keff), diferenciais (espectro, fluxo) e de queima e transmutação (inventário ao final do ciclo) e (iv) aplicar a metodologia desenvolvida em um estudo que possa ajudar na escolha futura de um sistema transmutador dedicado. Foram utilizados para tanto os seguintes códigos: MCNP (Transporte de partículas por Monte Carlo), MCB (acoplamento do MCNP com código de transmutação) e o sistema NJOY para o processamento dos arquivos de dados nucleares avaliados. / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

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Made available in DSpace on 2014-10-09T12:27:28Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:06:47Z (GMT). No. of bitstreams: 0 / Fundação de Amparo à Pesquisa do Estado de São Paulo (FAPESP) / Neste trabalho foram desenvolvidos calculos neutrônicos, termo-hidráulicos e de segurança para avaliar a seguranca operacional de um dispositivo de irradiação a ser colocado no núcleo do reator IEA-R1 do IPEN-CNEN/SP. Este dispositivo de irradiação é utilizado para alojar miniplacas de combustvel do tipo dispers~ao de U3O8-Al e U3Si2-Al, com 19,75% em peso de 235U e densidades, respectivamente, de ate 3,2 gU/cm3 e 4,8 gU/cm3. Estas miniplacas serão irradiadas a queimas acima de 50% do 235U, de forma a qualificar este tipo de dispersão para utilização no Reator Multipropósito Brasileiro (RMB), em concepção. Para os calculos neutrônicos, foram utilizados os programas computacionais 2DB e CITATION. O programa FLOW foi utilizado para determinar o fluxo de refrigerante no irradiador, permitindo o cálculo das temperaturas máximas atingidas nas miniplacas de combustível com o programa MTRCR-IEA-R1. Um Acidente de Perda de Refrigerante (APR) foi analisado com os programas computacionais LOSS e TEMPLOCA, permitindo o cálculo das temperaturas nas miniplacas de combustível após o esvaziamento da piscina do reator. Os cálculos demonstraram que a irradiação deverá ocorrer sem consequências adversas no núcleo de reator IEA-R1. / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP / FAPESP:08/55686-6