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Rod bundle is a typical constitutive element of a very wide range of nuclear reactor designs. This thesis describes the investigation of such geometry with wall-resolved Large Eddy Simulation (LES). In order to alleviate the mesh constraint, imposed by the near wall resolution, the usage of embedded refinements and polyhedral meshes is analysed firstly with a inviscid laminar case (Taylor Green vortices) and secondly with a fully turbulent case (channel flow only with embedded refinement). The inviscid test case shows that the addition of embedded refinements decreases the conservation properties of the code. Indeed the accuracy decreases from second order in a structured conformal mesh, to something in between first and second order depending on the quality of the unstructured mesh. Better results are obtained when the interface between refined and coarse areas presents a more regular and structured pattern, reducing the generation of skewed and stretched cells. The channel flow simulation shows that the Reynolds stresses, of some embedded refined meshes, are affected by spurious oscillations. Surprisingly this effect is present in the unstructured meshes with the best orthogonal properties. Indeed analysis of Reynolds stress budgets shows that terms, where the gradient in the wall normal direction is dominant, have a largely oscillatory behaviour. The cause of the problem is attributed to the convective term and in particular in the method used for the gradient reconstruction. As a consequence of these contradictory signs between the inviscid and the fully turbulent cases, the rod bundle test case is analysed using a conventional body fitted multiblock mesh. Two different Reynolds numbers are investigated reporting Reynolds stresses and budgets. The flow is characterised by an energetic and almost periodic azimuthal flow pulsation in the gap region between adjacent sub-channels, which makes turbulent quantities largely different from those in plane channel and pipes and enhances mixing. Experiments found that a constant Strouhal number, with the variation of the Reynolds number, characterises the phenomenon. The frequency analysis finds that present simulations are distinguished by three dominant frequencies, the first in agreement with the experimental value and two higher ones, which might be due to the correlation of the azimuthal velocity in the streamwise direction. Several passive temperature fields are added at the simulations in order to study the effects of the variation of the Prandtl number and the change in boundary conditions (Neumann and Dirichlet). A simplified case where an imbalance of the scalar between adjacent sub-channels is also investigated in order to evaluate the variation of the heat fluxes with respect to the homogeneous case. An alternative solution, to reduce the mesh constraint imposed by the wall, is to hybridize LES with RANS. The main achievement of this work is to integrate the heat transfer modelling to the already existing model for the dynamic part. Further investigations of the blending function, used to merge the two velocity fields, are carried out in conjunction with a study of the model dependency on the mesh resolution. The validation is performed on a fully developed channel flow at different Reynolds numbers and with constant wall heat flux. On coarse meshes the model shows an improvement of the results for both thermal and hydraulic parts with respect to a standard LES. On refined meshes, suitable for wall-resolved LES, the model suffers from a problem of double counting of modelled Reynolds stresses and heat fluxes because the RANS contribution does not naturally disappear as the mesh resolution increases.

Page iii was not included in the thesis. / Abstract Not Provided. / Thesis / Master of Engineering (MEngr)

Submitted by Almir Azevedo (barbio1313@gmail.com) on 2014-05-09T16:42:31Z No. of bitstreams: 0 / Made available in DSpace on 2014-05-09T16:42:31Z (GMT). No. of bitstreams: 0 Previous issue date: 2013 / Empregando a equação de difusão de nêutrons em estado estacionário multidimensional para simular o fluxo de nêutrons em reatores refrigerados água, e fazendo uso do Método de Diferenças Finitas, o presente trabalho tem por objetivo apresentar um estudo sobre o comportamento da distribuição de potência num reator tipo PWR, considerando a intensidade e a migração dos picos de potência à medida que ocorre a inserção das barras de controle no núcleo. Além disso, tomando como ponto de partida a distribuição axial de potência ao longo da vareta de maior fluxo de calor, realiza-se a análise térmica dessa vareta e do canal refrigerante associado. Para tal é empregado o código Fuel_Rod_3D, que usa o Método dos Elementos Finitos para modelar uma vareta combustível e seu canal refrigerante, possibilitando a simulação do comportamento termohidráulico de uma única vareta discretizada em três dimensões, considerando o fluxo de calor a partir do interior da pastilha combustível, passando pelo "gap" e pelo revestimento até alcançar o fluido refrigerante / This Work aims to present a study about the power distribution behavior in a PWR type reactor, considering both intensity and migration of power peaks due to insertion of control rods into the core. Employing the multidimensional steady-state neutron diffusion equation in order to simulate the neutron flux, and using the Finite Difference Method. Furthermore, based on the axial power distribution on the largest heat flux rod, is carried out thermal analysis of this rod and associated coolant channel. For this purpose is employed the Fuel_Rod_3D code, it uses the Finite Element Method to model the fuel rod and the associated coolant channel, allowing the thermohydraulics simulation of a single rod discretized in three dimensions, considering the heat flux from the pellet, crossing the gap and the cladding until it reaches the coolant.

Submitted by Almir Azevedo (barbio1313@gmail.com) on 2014-05-22T16:36:29Z No. of bitstreams: 1 dissertacao_mestrado_ien_2013_04 Renato Raoni.pdf: 2604427 bytes, checksum: ad33731b6c51a6c4e9e1abb26b746f83 (MD5) / Made available in DSpace on 2014-05-22T16:36:29Z (GMT). No. of bitstreams: 1 dissertacao_mestrado_ien_2013_04 Renato Raoni.pdf: 2604427 bytes, checksum: ad33731b6c51a6c4e9e1abb26b746f83 (MD5) Previous issue date: 2013-03 / Estudos de transferência de calor em varetas combustível são de grande importância na indústria nuclear. Isso se evidencia pela necessidade da predição de temperaturas limite para otimizar o projeto de varetas combustível. O presente trabalho tem por objetivo o desenvolvimento de um código computacional em linguagem fortran, no qual estão reunidas ferramentas como o Método de Elementos Finitos. Equações e correlações termo-hidráulicas foram implementadas no código com vistas a uma investigação profunda da transferência de calor entre a vareta combustível e o canal refrigerante, buscando, assim, entender o comportamento de ambos em regime transiente (como, por exemplo, nas situações de acidente). Foi feita uma análise sobre a validade da aproximação que desconsidera o fluxo axial de calor nas soluções analíticas. Comparações entre as soluções utilizando as propriedades constantes e propriedades dependentes da temperatura forma contempladas nesse trabalho. Estudos transientes envolvendo o desligamento do reator, considerando aspectos como a temperatura limite do combustível e o Departure from Nucleate Boiling Ratio (BNBR), foram realizados produzindo resultados que podem demonstrar o desempenho do código computacional / Studies on heat transfer fuel rods are of great importance in the nuclear industry . This is evidenced by the need for prediction of temperature limits to optimize the design of fuel rods . This work aims to develop a computer code in FORTRAN language, in which are gathered tools such as Finite Element Method . Equations and thermo- hydraulic correlations were implemented in the code with a view to a thorough investigation of heat transfer between the fuel rod and coolant channel , seeking thereby to understand the behavior of both in transient ( eg , in cases of accident ) . An analysis of the validity of the approach that disregards the axial heat flow in the analytical solutions was taken . Comparisons between solutions using constant properties and temperature dependent properties manner contemplated in this work . Transient studies involving the reactor shutdown , considering aspects such as limit the fuel temperature and Departure from nucleate Boiling Ratio ( BnBr ) , were performed yielding results that can demonstrate the performance of the computer code

京都大学 / 0048 / 新制・課程博士 / 博士(工学) / 甲第18589号 / 工博第3950号 / 新制||工||1607(附属図書館) / 31489 / 京都大学大学院工学研究科原子核工学専攻 / (主査)教授 功刀 資彰, 教授 中部 主敬, 講師 河原 全作 / 学位規則第4条第1項該当 / Doctor of Philosophy (Engineering) / Kyoto University / DFAM

京都大学 / 0048 / 新制・課程博士 / 博士(工学) / 甲第19994号 / 工博第4238号 / 新制||工||1656(附属図書館) / 33090 / 京都大学大学院工学研究科原子核工学専攻 / (主査)教授 功刀 資彰, 教授 佐々木 隆之, 准教授 横峯 健彦 / 学位規則第4条第1項該当 / Doctor of Philosophy (Engineering) / Kyoto University / DFAM

DYN3D is an best estimate advanced code for the three-dimensional simulation of steady-states and transients in light water reactor cores with quadratic and hexagonal fuel assemblies. Burnup and poison-dynamic calculations can be performed. For the investigation of wide range transients, DYN3D is coupled with system codes as ATHLET and RELAP5. The neutron kinetic model is based on the solution of the three-dimensional two-group neutron diffusion equation by nodal expansion methods. The thermal-hydraulics comprises a one- or two-phase coolant flow model on the basis of four differential balance equations for mass, energy and momentum of the two-phase mixture and the mass balance for the vapour phase. Various cross section libraries are linked with DYN3D. Systematic code validation is performed by FZR and independent organizations.

DYN3D is an best estimate advanced code for the three-dimensional simulation of steady-states and transients in light water reactor cores with quadratic and hexagonal fuel assemblies. Burnup and poison-dynamic calculations can be performed. For the investigation of wide range transients, DYN3D is coupled with system codes as ATHLET and RELAP5. The neutron kinetic model is based on the solution of the three-dimensional two-group neutron diffusion equation by nodal expansion methods. The thermal-hydraulics comprises a one- or two-phase coolant flow model on the basis of four differential balance equations for mass, energy and momentum of the two-phase mixture and the mass balance for the vapour phase. Various cross section libraries are linked with DYN3D. Systematic code validation is performed by FZR and independent organizations.

In recent years, the enthalpy raise hot channel factor limit has decreased significantly due to the power upgrade of Ringhals 4 and the use of shielding fuel assemblies. The shielding fuel assemblies task are to reduce the neutron leakage to the reactor vessel and in that way extend the reactor lifetime. This is achieved by replacing a few fuel rods with steel rods. Experiences from the last fuel cycles show that the core design procedure has been hampered because of this and that it’s hard to stay under the design limit. A way to overcome this problem and to improve the fuel economy is to introduce the use of radial enrichment distribution in the fuel assembly. This master thesis shows, through a case study of three fuel cycles at Ringhals 4,  that the internal peaking factor can be improved by roughly 2–3 % and that the maximum enthalpy raise hot channel factor can be improved by about 2.0–2.5 % if the fuel assemblies contain three different levels of enrichments instead of currently one. This can be achieved without any noticeable decrease in cycle length. / Genom en fallstudie av tre driftcykler för Ringhals 4 visar den här rapporten att max FΔH under cykeln kan sänkas med 2,0–2,5 % om bränsleknippena radiellt anrikningsoptimeras med minst tre delanrikningar. Totalt under cykeln kan FΔH sänkas med upp till 4 %. Om radiell anrikningsoptimering införs för Vattenfalls PWR:er skulle arbetet med att designa härdarna förenklas och utrymme ges för att ladda reaktorerna på ett mer ekonomiskt sätt.