A steady-state, single phase subchannel solver was created for the purpose of integration into a multi-physics nuclear fuel performance code. Since applications of such a code include full nuclear reactor core flow simulation, a thorough investigation of efficient solution techniques is a requirement. Execution time profiling found that formation of the Jacobian matrix required by the nonlinear Newton solve was found to be the most time-consuming step in solution of the subchannel equations, so several techniques were tested to minimize the time spent on this task, such as finite difference and the formation of an approximate Jacobian. Simple Jacobian lagging was shown to be very effective at reducing the total time computing the Jacobian throughout the Newton iteration process. Various linear solution techniques were investigated with the subchannel equations, such as the generalized minimal residual method (GMRES) and the aggregation- based algebraic multigrid method (AGMG). A number of physics-based preconditioners were created, based on a simplified formulation with no crossflow between subchannels, and it was found that of the preconditioners developed for this research, the most promising was a preconditioner that fully decoupled the subchannels by ignoring crossflow, conduction, and turbulent momentum exchange between subchannels. This independence between subchannels makes the task of parallelization in the preconditioner to be very feasible. However, AGMG clearly proved to be the most efficient linear solution technique for the subchannel equations, solving the linear systems in less than 5 percent of the time required for preconditioned GMRES.

The structure of turbulence in the subchannels of a large-scale 60 degree section of a CANDU 37-rod bundle was studied at Reynolds numbers equal to 50,000, 100,000 and 130,000. Measurements were conducted at roughly 33.81 rod diameters from the inlet of the rod bundle using single-point, two-component hot-wire anemometry. Analysis of the axial velocity signal indicated a weak effect of Reynolds number on the axial velocity distribution and a bulging of axial velocity contours toward the narrow gaps. The normalised normal Reynolds stresses and the normalised turbulent kinetic energy were found to decrease as the Reynolds number increased. The radial Reynolds shear stress varied linearly with radial distance from the rod, crossing zero at the location of local maximum of the axial velocity. This stress was symmetric about the central rod whereas the azimuthal Reynolds shear stress was anti-symmetric. The Reynolds number effect was weak but measurable on the integral length scales of the axial and radial velocity fluctuations but negligible on the integral length scale of the azimuthal velocity fluctuations, especially in the gap regions. The Taylor and Kolmogorov microscales increased from the wall toward the centre of the subchannel and decreased as the Reynolds number increased. The wall shear stress stress distribution around the central rod indicated no effect of Reynolds number, when normalized by the corresponding average. The wall shear stress reached local minima at rod-wall and rod-rod gaps and local maxima in the open flow regions. Vortex streets were generated within the subchannels very close to the inlet of the rod bundle. The convection speed and frequency of the vortex street were found to increase proportionately to Reynolds number, whereas the vortex spacing was not affected by the Reynolds number.

Orientadores: Roger Josef Zemp, Paulo de Carvalho Tofani / Tese (doutorado) - Universidade Estadual de Campinas, Faculdade de Engenharia Quimica / Made available in DSpace on 2018-08-03T15:26:11Z (GMT). No. of bitstreams: 1 Veloso_MarceloAntonio_D.pdf: 10344720 bytes, checksum: e173f03cd07f8090cdc60aacf528de23 (MD5) Previous issue date: 2003 / Resumo: O programa computacional P ANTERA-2 (Programa para Análise Termo-hidráulica de Reatores a Água, Versão 2), cujos fundamentos são descritos neste trabalho, efetua a análise por subcanais de feixes de varetas em conjunção com a simulação de múltiplos circuitos. O programa resolve simultaneamente as equações de conservação da massa, dos momentos axial e lateral e da energia para a geometria de subcanais acopladas com as equações de balanço que descrevem o escoamento de um fluido em um número arbitrário de circuitos de remgeração conectados a um vaso de pressão que contém o feixe. Atendo-se à formulação de subcanais, a estratégia computacional básica de P ANTERA-2 provém dos códigos COBRA, mas um método implícito alternativo de solução orientado para o campo de pressões é usado para resolver as aproximações de diferenças finitas das leis de balanço. Os resultados previstos pelo modelo de subcanais compreendem as distribuições de densidades, entalpias, vazões de massa e pressões nos subcanais. O modelo de circuitos prevê as vazões nos circuitos individuais, a vazão total através do vaso de pressão e as velocidades de rotação das bombas em função do tempo subseqüente à falha de qualquer número das bombas de circulação. Os transitórios de vazão nos circuitos podem ser ocasionados pelas perdas de potência elétrica, ruptura de eixos e travamento de rotores das bombas. As variações nas velocidades de rotação das bombas em função do tempo são determinadas através de um balanço de torques. A altura de recalque e o torque hidráulico das bombas são calculadas em função da velocidade e da vazão com duas curvas homólogas polares fornecidas ao programa na forma tabular. Para ilustrar a capacidade analítica de P ANTERA-2, três problemas-exemplo são apresentados e discutidos. Comparações entre resultados calculados e medidos indicam que o programa reproduz com boa precisão dados experimentais de temperaturas de saída de subcanais e de fluxos de calor críticos em feixes de 5x5 varetas. Observa-se támbém uma boa concordância entre as curvas teóricas previstas por P ANTERA-2 e valores medidos para as velocidades de rotação das bombas e vazões de massa nos circuitos primários da central nuclear Angra-2, quando suas quatro bombas principais são simultaneamente desligadas para simular o evento de declínio de vazão. Palavras-chave: análise por subcanais, código de subcanais, códigos cobra, análise de circuitos de escoamento, acidente de falha de bombas / Abstract: PANTERA-2 (from Programa para Análise Termo-hidráulica de Reatores a ÁguaProgram for Thermal-hydraulic Analysis of Water Reactors, Version 2), whose fundamentals are described in this work, is intended to carry out rod bundle subchannel analysis in conjunction with multiloop simulation. It solves simultaneously the conservation equations of mass, axial and lateral momentum, and energy for subchannel geometry coupled with the balance equations that describe the fluid flows in any number of coolant loops connected to A pressure vessel containing the rod bundle. As far as subchannel analysis is concemed, the basic computational strategy of P ANTERA-2 comes from COBRA codes, but an altemative implicit solution method oriented to the pressure field has been used to solve the finitedifference approximations for the balance laws. The results provided by the subchannel mode1 comprise the fluid density, enthalpy, flow rate, and pressure fields in the subchannels. The loop model predicts the individualloop flows, total flow through the pressure vessel, and pump rotational speeds as a function of time subsequent to the failure of any number of the coolant pumps. The flow transients in the loops may initiated by partial, total or sequentialloss of electric power to the operating pumps. Transient events caused by either shaft break or rotor locking may also be simulated. The changes in rotational speed of the pumps as a function of time are determined from a torque balance. Pump dynamic head and hydraulic torque are calculated as a function of rotational speed and volumetric flow from two polar homologous curves supplied to the code in the tabular form In order to illustrate the analytical capability of P ANTERA-2, three sample problems are presented and discussed. Comparisons between calculated and measured results indicate that the program reproduces with a good accuracy experimental data for subchannel exit temperatures and critical heat fluxes in 5x5 rod bundles. It is also observed a good correspondence between the theoretical curves predicted by P ANTERA-2 and measured values for pump rotational speeds and mass flow rates in the primary loops of Angra-2 nuclear power plant, when the four main coolant pumps are simultaneously switched off to simulate the flow decline evento / Doutorado / Sistemas de Processos Quimicos e Informatica / Doutor em Engenharia Química

Film flow measurements at several axial positions in round pipes with variousaxial power distributions are presented for conditions corresponding to normaloperation of a BWR. It is confirmed that the film flow rate approaches zero atthe onset of dryout. Selected phenomenological models of annular two-phaseflow are shown to reasonably reproduce the measurements. It is concluded thatmodels are in better agreement with measurements if terms corresponding topossible boiling induced entrainment are excluded. A method to perform film flow analysis in subchannels as a post-processto a standard two-field subchannel code is suggested. It is shown that thisapproach may yield accurate prediction of dryout power in rod bundles to alow computational cost and that the influence of the power distribution is wellpredicted by the model. / QC 20100618

Subchannel allocation schemes for OFDMA (orthogonal frequency division multiple access) systems with adaptive transmission are investigated. The analysis involves two scenarios: single-hop transmission and dual-hop relaying transmission. The overall objective is to utilize minimum system resource (in terms of subchannel) while satisfying scheduled users' data rate requirements under a certain error rate performance restriction. The ordered subchannel selection with adaptive modulation scheme is applied to both cases. In single-hop case, a low-complexity greedy subchannel allocation algorithm with two options is proposed, depending on whether a common modulation mode is applied on all selected subchannels or not. The resulting two options are termed as GS-FA (greedy selection with full adaptivity) and GS-LA (greedy selection with limited adaptivity), respectively. In dual-hop relaying case, two relaying schemes are considered: Amplified-and-Forward (AF) and Decode-and-Forward (DF). Based on this, four different combinations of relaying schemes and ordered subchannel selection with adaptive modulation scheme are investigated: DF with full adaptivity (DF-FA), AF with full adaptivity (AF-FA), AF with limited adaptivity (AF-LA), and AF with limited adaptivity and subchannel mapping (AF-LA with SCM). For both single-hop and dual-hop relaying case, performances of different proposed schemes are evaluated through mathematical analysis and compared with selected numerical results.

This master’s thesis deals with boiling crisis and with departure from nucleate boiling ratio. This thesis explains terms like the boiling crisis in nuclear reactors and the thesis deals with individual parameters of the reactor core, which have an impact on departure from nucleate boiling ratio. After that, the thesis deals with subchannel analysis and describes basic mathematical and physical models of the chosen subchannel program. The thesis then processes, with the ALTHAMC12 subchannel program, the chosen parameters and their impact on departure from nucleate boiling ratio. The conclusion of the diploma thesis deals with the evaluation of the best and worst calculated combination.

<p>Flow pulsations in subchannel geometries play an important role in homogenization of fluid temperatures within a fuel rod bundle cross-section. As such, there is a strong need to develop accurate integral models that incorporate the underlying physics of these flows for inclusion in the broader safety analysis codes. This research is concerned with using computational fluid dynamics to investigate the flow pulsations in order to develop an enhanced understanding of the flow physics. The vast majority of previous experimental work has been in the turbulent regime, with varying degrees of geometric complexity. Previous numerical work has focused on steady or unsteady simulation of the turbulent experimental results, with the requirement that an appropriate turbulence model must be selected.</p> <p>Recent experimental work by Gosset and Tavoularis in 2006 has indicated that flow pulsations can occur under laminar conditions. Computational modeling of laminar flow pulsations provides an ideal framework for studying the physical mechanisms or instabilities that promote formation of the pulsations. Simulations of their experimental domain were run for a gap height normalized by the rod diameter (δ/D) of 0.3 and Reynolds numbers of 718, 900 and 955. These simulations found frequencies in the same range as Gosset and Tavoularis, as well as qualitatively similar particle tracks to their dye streaks. Analysis of the numerical pulsations showed them to be fluid rotations around the rod. These rotations were shown to be strongly correlated with the axial velocity gradient, which acted to transfer momentum from axial flow to the crossplane rotational pulsatile flow. The pulsations were shown to develop from a purely axial flow through disturbances in the axial velocity gradient, which initially arose near inflection points in the axial velocity profile in the spanwise direction. Under the influence of the axial velocity gradient and fluctuating pressure, these disturbances evolve into a sustained quasi-periodic flow.</p> / Master of Applied Science (MASc)

The effects of appendages on flow characteristics and scalar mixing in gap-connected twin-subchannel geometries has been assessed. The assessment considers a symmetric, rectangular compound channel geometry connected by a single rectangular gap using computational fluid dynamics (CFD). Detailed numerical models (geometry and turbulence), characterizing the full test section from a reference experimental study, are generated and validated against measurements. Time varying details of the gap induced periodic structures and appendage induced vortices are captured through calculations in an unsteady Reynolds averaged Navier-Stokes (RANS) framework coupled with the Spalart-Allmaras (SA) turbulence model closing the RANS equations. Companion simulations are performed at each of two Reynolds numbers (2690 and 7500), one with and one without a gap-centered appendage. The appendage size modelled is representative of CANDU endplates. The appendage effects on flow characteristics and mixing are isolated through comparison of the associated simulations. In the absence of appendages, fluid exchange between subchannels is dominated by quasi-periodic flow pulsations through the gap formed due to flow instability in the near gap region. Without a gap-centered appendage, the magnitude, frequency, and structure length of the gap flow pulsations are well predicted by the model at both Reynolds numbers. The total tracer transfer between subchannels is reasonably well predicted for Re = 2690 (within approximately 17% of the experimental value). The model fails to capture the measured increase in scalar transfer through the gap with increased Reynolds number, underpredicting scalar mixing by 55% at Re = 7500. An argument is presented that the use of an isotropic turbulence model in the channel (SA), which precludes the development of channel secondary flows, is the source of the discrepancy between modelled and measured mixing at Re = 7500. Appendages, such as those introduced by end plates or bearing pads in CANDU fuel bundles, augment the exchange process between subchannels. With an appendage representative of a CANDU fuel bundle endplate introduced into the gap region, crossflow velocity and frequency are predicted to increase immediately downstream of the appendage due to flow diversion and vortex shedding. The higher local frequency is shown to be consistent with the vortex shedding frequency calculated for a stationary rectangular cylinder at the gap conditions. Further downstream, gap induced instabilities begin to re-establish as the dominant contributor to crossflow pulsations although they are not fully recovered by the test section exit. Mixing is augmented more by the appendage with increasing Reynolds number for the range examined. / Thesis / Master of Applied Science (MASc) / The fuel bundle and pressure tube assembly in the core of a CANDU reactor forms an intricate web of subchannels of varying geometries with interconnecting gaps. Heat generated within the fuel bundles is removed by coolant flowing through the pressure tube and within the bundle subchannels. Although flow is nominally axial along the length of the rod bundles, coolant is free to move between subchannels through the gaps by a variety of mechanisms. Detailed fluid flow in these rod bundle geometries is a complex 3D phenomenon, strongly affected by fluid turbulence and flow instabilities associated with the subchannel geometry. This flow is investigated in the current study and extended to include the effect of appendages, which hold the fuel rods in place, to determine their impact on mixing along the length of the bundle. Particular applications of the results of this study are in the areas of nuclear reactor performance and safety. The extent of coolant exchange between subchannels affects the local subchannel flow and temperature and, as a result, local cooling at the fuel element surfaces. Fuel element cooling is a principal component of reactor analysis under both normal operating conditions and postulated accident scenarios.

Um dos parâmetros termo-hidráulicos de segurança mais importantes no projeto e operação de reatores a água pressurizada é o fluxo de calor crítico (FCC). O FCC ocorre quando se atinge uma região de instabilidade na mudança de mecanismo de transferência de calor de uma parede aquecida para um fluido, aumentado drasticamente a temperatura da parede. Transientes em um reator nuclear podem afetar a taxa de geração de calor ou a fluxo de refrigerante no núcleo, prejudicando a retirada de calor das varetas combustíveis. Conhecer o FCC nestas condições é essencial para evitar danos às varetas e, consequentemente, a liberação de material radioativo. O objetivo deste trabalho é analisar o FCC para o LABGENE (Laboratório de Geração Nucleoelétrica) por meio do planejamento experimental e da simulação de seções de teste em condições de operação utilizando o código COBRAIIIc/MIT-1 e a correlação EPRI para o FCC. Considerou-se primeiramente seções de teste 3×3 de dois tamanhos distintos e os resultados para 100 pontos experimentais foram mostrados por meio de superfícies de resposta, a fim de melhor visualizar e analisar o comportamento de FCC para cada condição. Dois pontos importantes são os valores máximo e mínimo do FCC encontrados. O valor máximo (1,038 MBtu/hr.ft2 ou 3,27 MW/m2) indica o fluxo de calor necessário para a realização dos experimentos e o mínimo (0,162 MBtu/hr.ft2 ou 0,51 MW/m2) indica a pior condição de operação, a qual estaria mais próxima do ponto de ebulição. As simulações e modificações no código foram verificadas utilizando o banco de dados da Universidade de Columbia. Foram selecionados 2718 pontos experimentais referentes a seções de teste 5×5 com perfil de potência uniforme. Os resultados foram apresentados pela razão entre o valor predito e o valor experimental (DNBR) e os limites de tolerância unilateral 95/95 foram calculados, estando dentro dos valores esperados. / One of the most important thermal-hydraulic safety parameters for pressurized water reactor design and operation is the critical heat flux (CHF). The CHF occurs when a region of instability reached in the change of heat transfer mechanism from a hot wall to a fluid is reached, dramatically increasing the wall temperature. Transients in a nuclear reactor can affect the heat generation rate or the coolant flow in the core, impairing the removal of heat from the fuel rods. Knowledge of the CHF on these conditions is essential to prevent fuel rod damages and therefore the release of radioactive material. The main goal of this work is to analyze the CHF for LABGENE (Nuclear-electrical Generation Laboratory) by an experimental design and test sections simulation in operating conditions by using COBRAIIIc/MIT-1 code and the EPRI correlation for CHF. 3x3 test sections were initially considered for two different heights and outcomes for 100 experimental points were shown by means of response surfaces in order to better visualize and analyze the behavior of CHF for each condition. Two important points are the maximum and minimum values of the CHF found. The maximum value (1.038 MW/m2 or 3.27 MBtu/hr.ft2) indicates the power required for the experiments and the minimum one (0.162 MBtu/hr.ft2 or 0.51 MW/m2) indicates the worst operation condition, which would be closer to the boiling point. Code simulations and modifications were verified using the CHF database of Columbia University. 2718 data points pertaining to test sections 5×5 with uniform power profile were selected. The results were presented by the ratio between the predicted value and the experimental value (DNBR) and the limits of unilateral tolerance 95/95 were calculated, being within the expected values.

Um dos parâmetros termo-hidráulicos de segurança mais importantes no projeto e operação de reatores a água pressurizada é o fluxo de calor crítico (FCC). O FCC ocorre quando se atinge uma região de instabilidade na mudança de mecanismo de transferência de calor de uma parede aquecida para um fluido, aumentado drasticamente a temperatura da parede. Transientes em um reator nuclear podem afetar a taxa de geração de calor ou a fluxo de refrigerante no núcleo, prejudicando a retirada de calor das varetas combustíveis. Conhecer o FCC nestas condições é essencial para evitar danos às varetas e, consequentemente, a liberação de material radioativo. O objetivo deste trabalho é analisar o FCC para o LABGENE (Laboratório de Geração Nucleoelétrica) por meio do planejamento experimental e da simulação de seções de teste em condições de operação utilizando o código COBRAIIIc/MIT-1 e a correlação EPRI para o FCC. Considerou-se primeiramente seções de teste 3×3 de dois tamanhos distintos e os resultados para 100 pontos experimentais foram mostrados por meio de superfícies de resposta, a fim de melhor visualizar e analisar o comportamento de FCC para cada condição. Dois pontos importantes são os valores máximo e mínimo do FCC encontrados. O valor máximo (1,038 MBtu/hr.ft2 ou 3,27 MW/m2) indica o fluxo de calor necessário para a realização dos experimentos e o mínimo (0,162 MBtu/hr.ft2 ou 0,51 MW/m2) indica a pior condição de operação, a qual estaria mais próxima do ponto de ebulição. As simulações e modificações no código foram verificadas utilizando o banco de dados da Universidade de Columbia. Foram selecionados 2718 pontos experimentais referentes a seções de teste 5×5 com perfil de potência uniforme. Os resultados foram apresentados pela razão entre o valor predito e o valor experimental (DNBR) e os limites de tolerância unilateral 95/95 foram calculados, estando dentro dos valores esperados. / One of the most important thermal-hydraulic safety parameters for pressurized water reactor design and operation is the critical heat flux (CHF). The CHF occurs when a region of instability reached in the change of heat transfer mechanism from a hot wall to a fluid is reached, dramatically increasing the wall temperature. Transients in a nuclear reactor can affect the heat generation rate or the coolant flow in the core, impairing the removal of heat from the fuel rods. Knowledge of the CHF on these conditions is essential to prevent fuel rod damages and therefore the release of radioactive material. The main goal of this work is to analyze the CHF for LABGENE (Nuclear-electrical Generation Laboratory) by an experimental design and test sections simulation in operating conditions by using COBRAIIIc/MIT-1 code and the EPRI correlation for CHF. 3x3 test sections were initially considered for two different heights and outcomes for 100 experimental points were shown by means of response surfaces in order to better visualize and analyze the behavior of CHF for each condition. Two important points are the maximum and minimum values of the CHF found. The maximum value (1.038 MW/m2 or 3.27 MBtu/hr.ft2) indicates the power required for the experiments and the minimum one (0.162 MBtu/hr.ft2 or 0.51 MW/m2) indicates the worst operation condition, which would be closer to the boiling point. Code simulations and modifications were verified using the CHF database of Columbia University. 2718 data points pertaining to test sections 5×5 with uniform power profile were selected. The results were presented by the ratio between the predicted value and the experimental value (DNBR) and the limits of unilateral tolerance 95/95 were calculated, being within the expected values.