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A new combination of materials with matched indices of refraction was used in average and statistical PIV analysis of a 5 × 5 rod bundle with prototypical pressurized water reactor (PWR) mixing vane spacer grids. The new combination of materials shows improved matching of refractive index which enabled detailed PIV measurement of subchannel flow, even when viewed through a relatively large number of material interfaces. Convergence of average full field PIV measurements at ∼ 27 , 000 Reynolds number was demonstrated as independent of measurement plane depth relative to camera view window. A method for average vector post-processing elimination is proposed using validation statistics from PIV. The proposed method shows potential for eliminating vectors in areas of poor-convergence for planes with little interfering geometry. A full uncertainty quantification was performed on the full field of each PIV measurement planes. Similar to convergence rates, uncertainty demonstrated independence from measurement plane depth. Additional material interface between camera and measurement plane does not appear to have affected instantaneous velocity measurements’ contribution to average vector fields. Symmetric measurement planes were compared between 2 rod pitch intervals of the spacer grid. Velocity and statistically derived quantities show good coherence between symmetric planes. Measurements near and even below the tips of the mixing vanes were collected and analyzed.

In this work, a single step of coupled calculations for a fuel rod of IPR-R1 TRIGA was performed. The used me-thodology allowed to simulate the fuel pin behavior in steady-state mode for different power levels. The aim of this paper is to present a practical approach to perform coupled calculations between neutronic (Monte Carlo) and thermal-hydraulic (CFD) codes. For this purpose, is necessary to evaluate the influence of the water thermal-physical properties temperature variations on keff parameter. Besides that, Serpent Nuclear Code was used for the neutronics evaluation, while OpenFOAM was used for thermal-hydraulics. OpenFOAM si- mula-tions were made by using a modified chtMultiRegionFoam solver, developed to read Serpent output correctly. The neutronic code was used without any modifications. The results shows that this coupled calculations were consistent and that leads to encouraging further methodology development and its use for full core simulation. Also, the results shows good agreement with calculations performed using other version of OpenFOAM and Milonga as neutronic code.

Spacer grids are one of main components of a Pressurized Water Reactor (PWR) fuel assembly. They are able to improve heat transfer from rod bundles to the water flow by increasing turbulence and mixture of this flow. On the other hand the pressure drop increases because spacer grids. Experimental and Computational Fluid Dynamics (CFD) analysis have been used to understand how spacer grids affect the water flow. This analysis is important to improve spacer grids thermal-hydraulic performance. This paper aims to investigate numerically and experimentally the water flow through PWR spacer grids. The numerical and experimental procedures have been developed for a 5x5 rod bundle with spacer grids at the Nuclear Technology Development Center (CDTN) in Belo Horizonte, Brazil. At CDTN, measurements of the velocity components are acquired with a 2D LDV (Laser Doppler Velocimetry) system and the numerical results are obtained using ANSYS CFX code. The measurements are obtained at one height downstream from a spacer grid and compared to CFD simulations for a flow rate at Reynolds number of 5.4x104 . Results show good agreement between both methodologies. The great repeatability and low experimental uncertainty evaluated (< 1.24%) in this work can be used to validate other CFD codes.

Evaluation of gadolinium enrichment in Gadolinia ( Gd 2 O 3 ) burnable poison fuel rods was performed for a model of PWR Angra-2 reactor fuel assembly. A parametric study of the effect of 155 Gd enrichment on neutronic performance was investigated during burn-up for a proposed very low gadolinium design (VLowGad-FA) having 1% of Gadolinia in weight fraction. Low fraction gadolinium designs enable the Angra-2 reactor to use its licensed limit of 235 U to its maximum potential, that is 4.25%, with better reactivity behavior for some cases. Thermal transport performance of irradiated fuel was followed through estimation of inter-assembly peak pin power during burn-up. Results showed that Gd enriched design would behave very similar to LowGad assembly in terms of controlling local core power. It was confirmed in the neutronic context that low gadolinium design provides fuel cycle economic benefits while supporting longer cycle operation. It was estimated savings of 18 effective full power days if LowGad (2.0% Gadolinia, 4.25% 235 U ) is used in place of refGad (7.0% Gadolinia, 2.90% 235 U ). Respect the proposed design, savings are up to 41 effective full power hours if the burnable poison fuel rod VLowGad (1% Gadolinia, 99% 155 Gd ) is used in place of the commercial LowGad. The break-even point for Gd-155 enrichment cost for VLow155Gd99 fuel assembly was calculated to be 371 US$/gram-Gd-155 based in year 2018 prices of MWe for Angra-2.

The structuring project of the Brazilian Multipurpose Reactor (RMB) is responsible for meeting the capacity to develop and test materials and nuclear fuel for the Brazilian Nuclear Program. An irradiation test device (Loop) capable of performing fuel test for power reactor rods is being conceived for RMB reflector. In this work neutronic calculations have been carried out in order to determine parameters to the Loop basic design. Safe position determined to be 52.7 cm displaced from the core, where linear heat rate (q’)

In this work, a comparison between the capabilities of a spallation source and a fusion source in the burnup of reprocessed fuels and in the regeneration of thorium was carried out. We have performed Monte Carlo simulations of a sub-critical core loaded with thorium and reprocessed fuel under irradiation of two different neutron sources. Two different fuels reprocessed by the GANEX technique were evaluated, one spiked with depleted uranium and the other one spiked with thorium. The aim is to investigate the neutronic behavior of thorium and reprocessed fuels under different neutron sources and analyze the fuel evolution during 10 years of burnup.

In this work, a fine-mesh 1:1 Computational Fluid Dynamics (CFD) – Monte Carlo Neutron Transport (MC) coupled calculations for a IPR-R1 TRIGA MARK I fuel pin was performed. The proposed methodology allowed the simulation of the fuel pin behavior in steady-state for different meshes and the evaluation of the discretization uncertainty. Comparison between mesh-based and Constructive Solid Geometry (CSG) MC simulations was performed. An extended Grid Convergence Index (GCI) method was applied to quantify numerical discretization uncertainty for k eff , temperature and power density profiles. Serpent Nuclear Code was used for neutronics simulations, while OpenFOAM was used for thermal–hydraulics calculations. The proposed coupling method was proved convergent. This coupling method did not produce a smooth power density profiles. In order to achieve high-quality profiles, neutronic noise reduction procedure was performed. The results were encouraging, nevertheless, the methodology should be employed on the study of problems with increased complexity for further assessment.

Pressurized water reactors (PWRs) are the most common types of electricity generating nuclear reactors. Within their cores, fuel rods generate heat by fission and the high-pressure water is used as a coolant. The heated water is then used as a heat source in a steam generator that boils water in the secondary loop. The steam is used to spin a turbine and generate electricity. Spacer grids are key components of a PWR’s core. Their objectives are to maintain the fuel rods at their positions and enhance the coolant mixing and heat exchange. Over the past decades, numerous experimental and numerical studies have been performed, to characterize the flow induced by different types of spacer grids. The recent advancements of the computational description of the above-mentioned flows, led to a shift to higher resolution models for turbulence like Large Eddy Simulations. Therefore, highly spatially resolved experimental data is needed for code validation. The comparison and usage of the data about the flow in rod bundles with spacer grids obtained by different research groups is challenging. One major reason comes from the fact that the majority of the spacer grids studied are proprietary, and their geometries cannot be disclosed. The data required for some purposes like numerical code validation do not necessarily have to originate from one specific type of spacer grid. It is important to provide a detailed overview of the geometry tested and the experimental conditions. Equally crucial is to possess high spatial resolution and well-quantified and low relative uncertainty. In this study, a new, non-proprietary spacer grid is designed and 3D-printed. The flow induced by the spacer grid is characterized in a 5 × 5 rod bundle facility with Matching of Index of Refraction (MIR) at a Reynolds number Re = 27 , 390 using particle image velocimetry techniques. A complete uncertainty analysis is performed accounting for a variety of uncertainty sources. A low relative uncertainty is obtained. Eight middle-of-subchannel planes are selected as the domain of interest for this paper. Highly spatially resolved statistical results for the velocities, root-mean-square (RMS) fluctuating velocities, and Reynolds stresses are obtained along the span-wise direction of the flow, starting at the edge of the spacer grid up to 4.3 hydraulic diameters. The cross-flow between subchannels induced by the spacer grid is sustained for all the domain assessed. The turbulence induced by the spacer grid is assessed with the use of two-points cross-correlations and with the comparison of the RMS fluctuating velocities at different elevations. An increase of the size of the turbulent vortices along the span-wise direction is related to the decrease in the rms fluctuating velocities to evidence the turbulent dissipation. Differences between the size of the turbulent structures found with integral length scale calculations in the stream-wise and span-wise directions are attributed to the turbulence anisotropy. The full dataset and spacer grid geometry design will be available on the research group website of the authors. The data obtained can be used by other groups according to their interests.

Perforated plates with thin orifices is one of the options under considerations to be used in the debris filtering bottom end pieces for the next generation of fuel assemblies for the Brazilian nuclear plants. The increase in pressure drop can be drastically reduced by optimizing the chamfer geometries. This paper describes the development and validation of a procedure devised to use Computational Fluid Dynamics (CFD) to accurately estimate the pressure drop of water flow through perforated plates with chamfered thin orifices. The procedure comprises a scaled down calculation domain, grid configuration and a set of numerical solution parameters. The validation was performed against experiments with a set of plates with different chamfer geometric features like position (inlet, outlet and both sides), angles and sizes. Three turbulence models were evaluated with the standard k–ɛ giving the best result. For the range of parameters evaluated here the pressure drop decrease of a two sided chamfer plate can be estimated as a sum of both inlet and outlet chamfers individually. In both cases the pressure drop decreases rapidly for small chamfers and more slowly for larger chamfers. In the flat regions the chamfer angle has little influence at the inlet and is more important at outlet

Thermal stratification and striping are observed in many piping systems including those of nuclear power plants. Periodic occurrences of these thermal transients lead to fatigue and may induce undesirable failures and deformations to the piping. The Thermal Hydraulic Laboratory of the Centro de Desenvolvimento da Tecnologia Nuclear/Comissao Nacional de Energia Nuclear (CDTN/CNEN) conducts an experimental and numerical project simulating the thermal stratified flows in piping systems of pressurized water reactors (PWR) to obtain some understanding on these phenomena. Experiments were carried out in a test section simulating the steam generator injection nozzle of a PWR. A numerical simulation of one experiment was performed with the commercial finite volume Computational Fluid Dynamic code CFX 13.0. A vertical symmetry plane along the pipe was adopted to reduce the geometry in one half, reducing mesh element size and minimizing processing time. The RANS two equations RNG k–ɛ turbulence model with scalable wall function and the full buoyancy model were used in the simulation. In order to properly evaluate the numerical model it was performed a Verification and Validation (V&V) process. Numerical uncertainties due to mesh refinement and time step were evaluated. This validation process showed the great importance of a proper quantitative evaluation of numerical results. In past studies qualitative evaluations were considered enough and numerical results like the one presented here could be considered satisfactory for the prediction of thermal stratified flows. However, with the present V&V study it was possible to identify objectively the strengths and weaknesses of the model.

We evaluate the Reynolds-averaged Navier–Stokes equations available as commercial computational fluid dynamics code for the simulation of a neutral atmospheric boundary layer and attempt to define a proper numerical simulation procedure. Four turbulence models, including two-equation and Reynolds stress models, were evaluated together with two near-wall models. Mesh and map digitization sensitivity tests were also performed. The simulations were compared to experimental field data from the Askervein Hill in Scotland. The results show that the simulations performed with ANSYS CFX 12.1 on a proper mesh and topological map with a Reynolds stress turbulence model provided the best wind-speed predictions when compared to the experimental results.

The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due to the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5 ?? 5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 sub-domains. The sub-domains were simulated sequentially applying the outlet results of a previous sub-domain as inlet condition for the next. In this study the k ??? ?? turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and In et al. in one subchannel computational domains. Comparison between numerical and experimental results of lateral and axial velocities along of the rod bundle show good agreement for all evaluated heights downstream the spacer grid. The present numerical procedure shows better predictions than Karoutas et al. model especially further from the spacer grid where the peripheral subchannels have more influence in the average flow.

An experimental and numerical study was conducted on the pressure loss of flows through the bottom end piece of a nuclear fuel assembly. To determine an optimized numerical methodology using the commercial CFD code, CFX 10.0, a series of preliminary simulations of water flows through perforated plates in a square ducts were performed. A perforated plate is a predominant geometry of the bottom end piece, responsible for the majority of the flow’s pressure drop. The numerical pressure loss applying an optimized mesh and the k-e turbulence model showed good agreement when compared with a conventional methodology (Idelchik). Numerical results for the standard bottom end piece were obtained applying the previously determined mesh criteria and the k-e turbulence model with some geometric simplifications. The agreement between the numerical simulations and experimental results can be considered satisfactory but suggests further numerical investigations with the bottom piece under real conditions of the experiment, without the geometric simplifications and with a gap between the piece and the wall of the flow channel. Additionally, other turbulence models should be appraised for this complex geometry.

Os elementos combustíveis nucleares de reatores de água pressurizada (PWR Pressurized Water Reactor) consistem de feixes de varetas combustíveis mantidas em arranjo quadrado por grades espaçadoras posicionadas ao longo de seu comprimento. As grades espaçadoras provocam fluxos transversais entre os subcanais e/ou rotacional no interior do próprio subcanal, aumentando a eficiência de troca térmica nas imediações da grade. Devido à discretização detalhada do domínio fluido, investigações numéricas utilizando a Dinâmica dos Fluidos Computacional (CFD Computational Fluid Dynamics) têm sido realizadas para o estudo desta importante característica termo-hidráulica e estão se tornando rapidamente aferramenta preferida para o desenvolvimento de grades espaçadoras. No entanto, antes que ferramentas CFD possam ser consideradas confiáveis, há a necessidade de estabelecer a credibilidade dos resultados numéricos. Procedimentos devem ser definidos para avaliar o erro e incertezas do modelo numérico. Estes procedimentos são referidos como processos de Verificação e Validação (V&V). Com o intuito de subsidiar o desenvolvimento da indústria nuclear no Brasil com ferramentas para o desenvolvimento de um elemento combustíveltotalmente nacional para reatores PWR, este trabalha visa o desenvolvimento de uma metodologia numérico-experimental para a investigação do escoamento de água através de grades espaçadoras com dispositivos de mistura. A metodologia desenvolvida empregou aferramenta de modelagem CFD e experimentos em um segmento do elemento combustível nuclear com 5 x 5 varetas para levantamento da queda de pressão e dos perfis de velocidade, obtidas por meio de um LDV (Laser Doppler Velocimetry), do escoamento após a grade espaçadora. Dados da literatura foram utilizados para o desenvolvimento inicial do procedimento de simulação numérica. Um processo de V&V foi realizado de acordo com a norma ASME V&V 20. A incerteza numérica foi estimada e cinco modelos de turbulência foram avaliados. O procedimento numérico verificado e validado foi aplicado na simulação da grade espaçadora testada experimentalmente. Os resultados numéricos foram comparados aos dados experimentais obtidos neste trabalho. Os resultados numéricos mostraram boa concordância qualitativa com os dados experimentais de diferença de pressão ao longo do feixe de varetas e perfis de velocidade. Os procedimentos apresentados neste trabalho podem ser úteis para um futuro desenvolvimento de uma grade espaçadora. The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The spacer grids cause a cross and swirl flow between and within the subchannels, enhancing heat transfer in the grid vicinity. Experimental and theoretical investigations, such asComputational Fluid Dynamics (CFD) analysis, have been carried out in the past years to study these important thermal and fluid dynamic features. Due to the detailed discretization of the flow characteristics obtained through CFD simulations, this analysis is quickly becoming the preferred tool for the improvement of spacer grid designs. However, before CFD can be considered as a reliable tool for the analysis of flow through rod bundles there is a need to establish the credibility of the numerical results. Procedures must be defined to evaluate the error and uncertainty due to aspects such as mesh refinement, turbulence model, wall treatment and appropriate definition of boundary conditions. These procedures are referred to as Verification and Validation (V&V) processes. With the intention to subsidize the development of the Brazilian nuclear industry with tools for the development of a totally national nuclear fuel assembly for PWR reactor, this work aims for the definition of an experimental and numerical methodology to investigate the water flow through spacer grids with mixing devices. The developed methodology employs the CFD tool and experiments in a 5 x 5 rod bundle fuel element segment to assess pressure loss and velocity profiles, obtained with an LDV (Laser Doppler Velocimetry) system. Data from literature was used for the initial development of the numerical procedure. A V&V process was performed according to the ASME V&V 20 standard. Numerical uncertainty was estimated and five turbulence models were evaluated. The verified and validated numerical procedure was applied to the simulation of the experimentally tested spacer grid. The numerical uncertainty values were extrapolated from the initial development. The numerical results were compared to the experimental data obtained in a test section developed in this work. The numerical results showed good agreement to the experimentally obtained data for pressure loss and the velocity profiles. The procedures presented and developed in this work can be very useful for future projects of spacer grids for nuclear fuel elements.

The spacer grids exert great influence on the thermal hydraulic performance of the PWR fuel assembly. The presence of the spacers has two antagonistic effects on the core: an increase of pressure drop due to constriction on the coolant flow area and increase of the local heat transfer downstream the grids caused by enhanced coolant mixing. The mixing vanes, present in most of the spacer grid designs, cause a cross and swirl flow between and in the subchannels, enhancing even more the local heat transfer at the cost of more pressure loss. Due to this important hydrodynamic feature the spacer grids are often improved aiming to obtain an optimal commitment between pressure drop and enhanced heat transfer. In the present work, the fluid dynamic performance downstream a 5 × 5 rod bundle with spacer grids is analyzed with a commercial CFD code (CFX 11.0). Eleven different split vane spacer grids with angles from 16° to 36° and a spacer without vanes were evaluated. The computational domain extends from ∼10 Dh upstream to ∼50 Dh downstream the spacer grids. The standard k-ε turbulence model with scalable wall functions and the total energy model were used in the simulations. The results show a considerable increase of the average Nusselt number and secondary mixing with the angle of the vane up to ∼20 Dh downstream the spacer, reducing greatly the influence of the vane angle beyond this region. As expected, the pressure loss through the spacer grid also showed considerable increase with the vane angle.

IRAN.LIB is a coupled neutron-gamma library which was developed to be used by the ANISN/PC (CCC-0514/02). The library was developed mainly for shielding calculations and it contains 33 elements mostly used as shielding materials, such as concrete. IRAN.LIB is a set of six libraries, each having the same elements but a different number of energy groups. In order to use IRAN.LIB by ANISN/PC, this code must be modified.

O programa comercial de CFD, o ANSYS CFX 10.0, foi utilizado para definir um procedimento numerico com a finalidade de avaliar a perda de carga em escoamento permanente, incompressivel e adiabatico atraves do Bocal inferior Padrao (BP) do elemento combustivel deum reator nuclear. Os resultados numericos foram comparados a resultados experimentais de perda de carga obtidos no Circuito Hidrodinamico do Reator (CHR) do Centro Tecnologico da Marinha (CTMSP). Inicialmente foram realizados estudos em placas perfuradas, geometria predominante no BP, aplicando-se simetria de 1/8 para determinar parametros otimos de malha e o modelo de turbulencia mais adequado. Os modelos de turbulencia RANS de duas equacoesk-Ã, k-Ö e SST foram avaliados. A influencia dos chanfros nos orificios da placa perfurada do BP, de uma folga entre o BP e o canal de escoamento presente nos experimentos, e da geometria que antecede a secao de testes tambem foram avaliadas. As simulacoesconfirmaram que a presenca dos chanfros provoca uma grande reducao da perda de carga (~40%) e do comprimento de recuperacao da pressao na placa perfurada. A folga provoca uma reducao de ~5% na perda de carga devido ao escoamento proximo da parede e ao aumento da secao transversal do canal. O aumento de malha impediu, devido a limitacoes computacionais, que a folga fosse simulada no BP, mas uma compensacao na vazao devido ao aumento da secao transversal foi aplicada. Foi constatado que a geometria que antecede a secao de testes promove uma nao uniformidade nos perfis de velocidade e de turbulencia que acarretam em reducao de ~ 4% na perda de carga atraves da placa perfurada. A presenca desta geometria foi omitida nomodelo final da simulacao do BP por impossibilitar que a simetria de 1/8 fosse aplicada, o que impediria a simulacao. O BP foi entao simulado em simetria de 1/8 com os modelos de turbulencia k-Ã e SST para todas as condicoes experimentais, com perfis uniforme de velocidade e turbulencia na entrada do canal. Os resultados de perda de carga obtidos com o modelo SST e k-Ã foram em media ~19% e ~10% maiores, respectivamente, do que os obtidos experimentalmente. As discrepancias observadas podem ser associadas principalmente assimplificacoes, realizadas devido a limitacoes computacionais, que mostraram uma tendencia de reducao na perda de carga calculada. Este trabalho devera subsidiar o desenvolvimento de um elemento combustivel nuclear avancado em curso na INB. The commercial CFD program, ANSYS CFX 10.0, was used to define a numerical procedure to evaluate the pressure drop in a permanent, incompressible and adiabatic flow through the standard Bottom end Piece (BP) of a nuclear fuel assembly. The numerical results were compared to experimental pressure loss results obtained at the Reactor Hydrodynamic Circuit (CHR) of the Centro Tecnológico da Marinha (CTMSP). Initial studies on 1/8 symmetry perforated plates, which are the predominant geometry of the BP, were performed to define the optimal mesh parameters and adequate turbulence model. The RANS two equations turbulence models k-e, k-. and SST were assessed. The influences of the chamfers of the BPs perforated plate orifices, of a gap between the BP and the duct present in the experiments, and of the geometry upstream of the test section were also appraised. The simulations confirmed that the presence of chamfers causes a drastic reduction of both pressure loss (~40%) and pressure recuperation length. The gap causes a reduction of ~5% of the pressure loss due toboth the flow near the wall and the increase of the cross section area of the duct. The increase in the mesh prevented, due to computational limitations, the simulation of the gap with the BP, but a compensation of the mass flow due to the increase of the cross section area was applied. It was observed that the geometry upstream of the test section promotes a nonuniform velocity and turbulence profiles that cause a ~4% reduction of the pressure loss at the perforated plate. The presence of this upstream geometry was omitted in the final BP simulation model for it disables the use of a 1/8 symmetry, which would prevent the simulation. Then a 1/8 symmetry of the BP was simulated with the turbulence models k-e and SST for all experimental conditions, applying a uniform profile of velocity and turbulence at the inlet. The pressure loss results obtained with the SST and k-e models were in average ~19% and ~10% higher, respectively, than the experiments. The discrepancies observed can be mainly associated to the simplifications, performed due to computational limitations, that shown a tendency to reduce the calculated pressure loss. This work shall support the development of a new advanced nuclear fuel element at INB.

Power monitoring of nuclear reactors is normally done by means of neutronic instruments, i.e. by the measurement of neutron flux. The greater the number of channels for power measuring the greater is the reliability and safety of reactor operations. The aim of this research is to develop new methodologies for on-line monitoring of nuclear reactor power using other reliable processes. One method uses the temperature difference between an instrumented fuel element and the pool water below the reactor core. Another method consists of the steady-state energy balance of the primary and secondary reactor cooling loops. A further method is the calorimetric procedure whereby a constant reactor power is monitored as a function of the temperature-rise rate and the system heat capacity. Another methodology, which does not employ thermal methods, is based on measurement of Cherenkov radiation produced within and around the core. The first three procedures, fuel temperature, energy balance and calorimetric, were implemented in the IPR-R1 TRIGA nuclear research reactor at Belo Horizonte (Brazil) and are the focus of the work described here. Knowledge of the reactor thermal power is very important for precise neutron flux and fuel element burnup calculations. The burnup is linearly dependent on the reactor thermal power and its accuracy is important in the determination of the mass of burned 235U, fission products, fuel element activity, decay heat power generation and radiotoxicity. The thermal balance method developed in this project is now the standard methodology used for IPR-R1 TRIGA reactor power calibration and the fuel temperature measuring is the most reliable way of on-line monitoring of the reactor power. This research project primarily aims at increasing the reliability and safety of nuclear reactors using alternative methods for power monitoring.

In order to study the safety aspects connected with the permanent increase of the maximum steady state power of the IPR-R1 TRIGA Reactor of the Nuclear Technology Development Centre (CDTN) at Belo Horizonte (Brazil), experimental measures were performed with the reactor working in power steps from ‘zero’ power up to 265 kW. The experiments were performed with the pool forced cooling system turned off. A number of parameters were measured in real-time such as fuel and water temperatures, radiation levels, reactivity and influence of cooling system. This paper summarises the behaviour of the operational parameters and presents some of the recent results obtained. Information on all aspects of reactor operation was displayed on the Data Acquisition System (DAS) shown the IPR-R1 real-time performance. The DAS was developed to monitor and record all operational parameters. Information displayed on the monitor was recorded on hard disk in a historical database.

This article presents a numerical investigation of thermal-fluid dynamics processes through the gaps of the spherical fuel elements (fuel pebbles) in the core of a Pebble Bed Reactor (PBR), using Computational Fluid Dynamics (CFD). The PBR is one of the most promising projects of the six classes of Generation IV Very High Temperature Reactor (VHTR). The results of two analyzes are presented. In the first case were evaluated two models of heat transfer to the spherical fuel. In this model is specified the volumetric heat generation, with thermal conduction in the fuel, and in the cladding. In the second model was specified a particular heat flux at the spherical fuel elements surface. In this analysis were performed simulations in two arrays of spheres i.e., the spheres into contact and spaced 2 mm. In the second analysis was also evaluated the influence of the spheres arrangement in the bed thermal-fluid dynamic behavior. The set of pebbles that constitute the core was modeled by representations of crystalline structure with different packing factors. The four simulations of the first analysis showed differences in flow and temperature profiles and maximum surface coating. There were also no significant differences in flow and heat transfer between the beads and the fluid in cases with different packing factors. These results show the importance of simulation of heat conduction inside the pebble fuel as well as the need to better assess the influence of the arrangement formed by pebbles fuel in PBR reactors thermal-fluid dynamics behavior.

O código STHIRP-1 foi desenvolvido como uma atividade da linha de pesquisa relacionada à área de termo-hidráulica de reatores, do Departamento de Engenharia Nuclear da Universidade Federal de Minas Gerais e é resultado de um esforço no sentido de desenvolver um código que tenha a mesma capacidade analítica daqueles desenvolvidos em instituições e centros de pesquisa qualificados na área nuclear em todo o mundo. A capacidade analítica do programa foi testada com a simulação do sistema representado pelo reator de pesquisa TRIGA IPR-R1 instalado no CDTN/CNEN em Belo Horizonte. Neste contexto resolveu-se avaliar o combustível anelar para testar o modelo de condução térmica implementado no programa. Para tanto simulou-se o combustível com geometria anelar apresentado no relatório de referência publicado pelo MIT Center for Advanced Nuclear Energy Systems em 2006. Nesse relatório discute-se a proposta de um combustível anelar, refrigerado interna e externamente, com o objetivo de permitir o aumento da densidade de potência de um reator PWR sem comprometer as margens de segurança da instalação. As condições termohidráulicas foram calculadas com o auxílio do código de subcanais VIPRE. Os cálculos foram realizados sempre procurando seguir as informações contidas no relatório de referência. Os resultados mostraram que, de uma forma geral, há uma boa correspondência entre aqueles previstos por STHIRP-1 e os apresentados no relatório. No modelo de uma vareta o bom acordo na comparação do fluxo de calor, do DNBR ( Departure From Nucleate Boiling) e queda de pressão do VIPRE e STHIRP confirmam a capacidade nativa de simulação do combustível anelar com o código STHIRP. Do modelo de 1/8 do núcleo, representando o núcleo por simetria com potência de 150% em relação ao uso do combustível sólido, foi obtido bom acordo do DNBR. Além disso, foram obtidas as distribuições de temperatura radial e axial da vareta combustível e a temperatura da água. A perda de carga foi o parâmetro, que apresentou maior diferença no trabalho. O canal interno apresentou a menor queda de pressão, pois não tem grade espaçadora. Enfim, este estudo mostra que o programa STHIRP é uma ferramenta valiosa que está à disposição do Departamento de Engenharia Nuclear sem custos e gera perspectivas de trabalhos futuros. O combustível anelar se mostra muito promissor, mas ainda é necessário mais estudos e investimento nesse combustível. The STHIRP-1 code was developed as an activity of the research line related to the thermal hydraulics area of reactors of the Department of Nuclear Engineering of Universidade Federal de Minas Gerais. Is the result of an effort to develop a code that has the same analytical capacity as those developed in institutions and research centers qualified in the nuclear area worldwide. The analytical capacity of the program was tested with the simulation of the system represented by the TRIGA IPR-R1 research reactor installed in the CDTN / CNEN in Belo Horizonte. In this context, it was decided to evaluate the annular fuel, to test the thermal conduction model implemented in the program. The fuel with annular geometry presented in the reference report published by the MIT Center for Advanced Nuclear Energy Systems in 2006 was simulated. This report discusses the proposal for an annular fuel, cooled internally and externally, in order to increase the power density of a PWR reactor without compromising the safety margins of the installation. The thermo-hydraulic conditions were calculated with the aid of the VIPRE. The calculations were always made to follow the information contained in the reference report. The results showed that, in general, there is a good correlation between those predicted by STHIRP-1 and those presented in the report. In the model of a dipstick the good agreement in the comparison of the heat flow, DNBR and pressure drop of VIPRE and STHIRP confirm the native simulation capacity of the annular fuel with the code STHIRP. From the 1/8 core model, representing the complete core by symmetry with a power of 150% in relation to the use of the solid fuel, good agreement of the DNBR (Departure From Nucleate Boiling), and heat flow was obtained. Besides, the radial temperature distributions were obtained and axial direction of the fuel rod and the water temperature. The load loss was the parameter, which showed the greatest difference, in the work. The internal channel had a lower pressure drop, because it does not have a spacer grid. Finally, this study shows that the STHIRP program is a valuable tool that is available to the Department of Nuclear Engineering at no cost and generates perspectives for future work. Ring fuel is very promising, but more research and investment is still needed.

O Reator Multipropósito Brasileiro (RMB) é um reator de pesquisa de 30 MW planejado para ser construído no Brasil. Neste tipo de reator, o núcleo é submerso em uma piscina profunda com água puricada e desmineralizada. A coluna de água acima do núcleo fornece blindagem radiológica para a sala do reator, permitindo também o acesso diretoao núcleo pelos operadores para manutenção e reabastecimento. Entretanto, pequenas concentrações de substâncias dissolvidas permanecem na água da piscina, sendo ativados quando submetidos ao uxo de nêutrons do núcleo. A água contendo esses elementos ativos se eleva para a superfície por convecção, elevando a taxa de dose a qual os trabalhadores são expostos na sala do reator. Em reatores de piscina aberta mais potentes, como o RMB, a taxa de dose pode ultrapassar os limites ocupacionais, sendo introduzido um circuito termo-hidráulico que cria uma Camada de Água Quente (CAQ) puricada no topo da piscina que limita a chegada de elementos ativos na superfície. O Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) construiu uma bancada experimental em escala 1/10 da piscina do RMB para o estudo do comportamento da CAQ e validação de modelos numéricos. Esse trabalho apresenta os resultados do estudo do impacto na estabilidade da CAQ pelo aquecimento da piscina devido à partida do reator. The Brazilian Multipurpose Reactor (RMB) is a 30 MW open pool research reactor planned to be constructed in Brazil. Such type of reactor is built inside a deep pool of puried and demineralized water, providing radiological protection still keeping the core accessible for maintenance and refueling. However, dissolved ions become activated in the pool water due to the core neutron ux, releasing radiation in the reactor room when the activated elements reach the top. Thus high power open pool reactors, as RMB, have an auxiliary thermal-hydraulic circuit that creates a Hot Water Layer (HWL) on the pools top, keeping the activated water under the HWL and mitigating the dose rate to which the operators are exposed to. The Centro de Desenvolmento da Tecnologia Nuclear (CDTN) built a 1/10 scale experimental bench of the RMBs pool for the HWL investigation. This work presents the results of the pools heating due to the reactor startup in the HWL stability.

Nesta tese descreve-se o desenvolvimento de um sistema de cálculos neutrônicos e termohidráulicos acoplados que utiliza malhas finas não-estruturadas como domínio de solução e totalmente baseado em software livre. As contribuições propostas seguem em dois diferenteseixos: uma delas é o foco na utilização da forma de desenvolvimento de sistemas de computação abertos, um conceito amplamente utilizado em diversas áreas do conhecimento mas raramente explorado no campo de Engenharia Nuclear; a segunda é o uso de memória compartilhada do sistema operacional como uma área de armazenamento de dados de confiável e rápido acesso para acoplar o sistema de dinâmica dos fluidos computacional (CFD) OpenFOAM e o gratuito e flexível código de análise de reatores milonga. Este conceito foi aplicado na simulação do comportamento de um modelo simplificado de um combustível do reator TRIGA Mark 1 IPR-R1 em estado estacionário. As seções de choque macroscópicas utilizadas pelo modelo, um conjunto para dois grupos de energia, foram geradas utilizando o código WIMSD-5B. Os resultados mostram que esta inovadoraforma de acoplamento neutrônico e termo-hidráulico leva a resultados consistentes, encorajando o futuro desenvolvimento deste sistema e seu uso na simulação de sistemasnucleares complexos. The development of a fine mesh coupled neutronics/thermal-hydraulics framework mainly using open source software is presented. The contributions proposed go in two different directions: one, is the focus on the open software development, a concept widely spread in many fields of knowledge but rarely explored in the nuclear engineering field; the second, is the use of operating system shared memory as a fast and reliable storage area to couple the computational fluid dynamics (CFD) software OpenFOAM to the free and flexible reactor core analysis code milonga. This concept was applied to simulate the behavior of the TRIGA Mark 1 IPR-R1 reactor fuel pin in steady-state mode. The macroscopiccross-sections for the model, a set of two-group cross-sections data, were generated using WIMSD-5B code. The results show that this innovative coupled system gives consistent results, encouraging system further development and its use for complex nuclear systems.

Os avanços computacionais observados nas duas últimas décadas têm proporcionado impacto direto nos trabalhos de simulação nuclear nos quais se utilizam uma multiplicidadede códigos computacionais, inclusive acoplados entre si, que possibilitam simulações analíticas e representam com notável precisão o comportamento real de instalações nucleares. Deste modo, estudos de cenários complexos em reatores nucleares têm sido aperfeiçoados pela utilização de sistemas de códigos termo-hidráulicos (TH) e neutrônicos (NK) acoplados. Essa técnica consiste em incorporar a modelagem neutrônica tridimensional nos códigos de sistemas para simular principalmente transitórios que envolvem distribuições assimétricas de potência e forte efeito de realimentação entre a neutrônica e a termo-hidráulica. Neste contexto, este trabalho apresenta uma série de etapas necessárias ao desenvolvimento de uma metodologia de acoplamento entre códigos de análise termo-hidráulica e códigos de análise neutrônica utilizando como modelo o reator de pesquisa TRIGA IPR-R1. Os resultados obtidos são satisfatórios e representam parte importante do desenvolvimento de tal metodologia. Os códigos envolvidos neste trabalho foram: RELAP5-MOD3.3, RELAP5-3D 3.0.0, PARCS 2.7 e WIMSD-5B além de várias interfaces gráficas e ferramentas computacionais acessórias The computational advances observed in the last two decades have provided direct impact on the researches related to nuclear simulations which use several types of computer codes, including coupled between them, allowing also the analytical simulation and represent with very much accuracy the real behavior of nuclear plants. In this way, studies of complex scenarios in nuclear reactors have been improved by the use of thermal-hydraulic (TH) and neutron kinetics (NK) coupled system codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into system codes, mainly to simulate transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronics and reactor thermal-hydraulics. Therefore, this work presents a series of steps needed to develop a proposal for a methodology of coupling between thermal-hydraulic and neutron kinetics system codes using as a model the research reactor TRIGA IPR-R1. The results obtained are satisfactory and represent important part of the development of this methodology. The codes involved in this work were: RELAP5-MOD3.3, RELAP5-3D 3.0.0, PARCS 2.7 and WIMSD-5B beyond of several graphical interfaces and computational accessory tools