Possui graduação em Engenharia Civil pela EPUSP (1977), mestrado em Tecnologia Nuclear pela Universidade de São Paulo (1980) e doutorado em Engenharia Civil - Estruturas pela EPUSP (1991). Atualmente é professor e orientador de Mestrado e Doutorado na área de Tecnologia Nuclear - Reatores da Universidade de São Paulo , consultor de empresas de projeto e fabricantes em projeto mecânico e estrutural de vasos de pressão e tubulações, e tecnologista senior da Comissão Nacional de Energia Nuclear. Tem mais de 30 anos de experiência na área de Engenharia Estrutural, com ênfase em avaliação estrutural e avaliação de integridade estrutural de componentes mecânicos, atuando principalmente nos seguintes temas: vasos de pressão, análise estrutural, método dos elementos finitos, integridade estrutural e análise de tensões. (Texto extraído do Currículo Lattes em 17 nov. 2021)

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Agora exibindo 1 - 10 de 239
  • Artigo IPEN-doc 30216
    Development of the reliability assurance program in a Brazilian nuclear power plant subsidized by a reliability, availability and maintainability model
    2023 - GOMES, J.M.; NETO, M.M.; MATURANA, M.C.; OLIVEIRA, P.S.P.
    The main objective of this work is to present a methodology for the development of a Reliability Assurance Program (RAP) specific to a PWR experimental nuclear installation, through the analysis of the installation and the development of a preliminary RAP subsidized by a Reliability, Availability and Maintainability (RAM) model. The study of an evaluation was carried out in the long-term decay heat removal of the studied experimental plant, whose data were used for application of the RAP. The necessary steps for applying the developed RAP are followed, using the data from the assessment of the studied plant, resulting in a list of components of significant risk for the Program, and in the following steps of sending the list to the experts panel, ranking of SSCs by the panel and development of the final list of significant risk SSCs for using the list in the optimization of the plant. The RAP subsidized by a RAM model will be able to work with the logical relationships between each component of the plant for their effects on energy generation and with the quantitative prediction of the magnitude of each contributor to the occurrence of high-level events, and the developed methodology can be applicable throughout the experimental plant. In this way, it will be possible to implement the RAP in the plant, which will provide a structured way to meet the regulatory requirements for its licensing. Also, it will be possible to complement the plant safety analysis report, which must contain the RAP.
  • Resumo IPEN-doc 30153
    Correlations of mechanical properties by SPT (Small Punch Test) and conventional tensile test for Al 6061 – T6
    The Small Punch Test (SPT) was development by nuclear industries to analyses mechanical properties of irradiated materials principally by small volume of the samples. This technique intend to evaluate the materials behavior during the time life of nuclear reactors, where yours properties changed by irradiation intensity and exposition time. It is considered an almost ¨non-destructive” method [2] due to small sample volume and its applications are spreading for use in situations where conventional methods do not apply. SPT consists of pressing a sphere, with a diameter equal to 2.5 mm, in a miniaturized sample of circular geometry (diameter d = 8 mm and thickness about 0.5 mm)[1], which has fixed edges, tested in conventional mechanical testing machines with the aid of a device developed for their achievement. In this work, mechanical properties of aluminum (Al 6061-T6) were abstained by two different methods: conventional tensile test and the small punch test (SPT). The SPT results depends on graph interpretations and discussions take place at now. Correlations of results guide us in choosing the most appropriated method for interpreting the force x displacement graph from SPT.
  • Artigo IPEN-doc 29947
    Crack tunneling effects on the elastic unloading compliance of C(T), SE(B) and clamped SE(T) specimens and correction methodology
    This paper covers the effects of crack tunneling on SE(B), C(T), and clamped SE(T) specimens and presents a correction methodology for this effect and is divided in two parts. Part one presents an investigation of how crack front curvature affects instantaneous crack size predictions based on the elastic unloading compliance technique. Relative crack depths (a/W) of 0.2, 0.5, and 0.7, were considered alongside five levels of crack curvature. Refined finite element models provided load-CMOD records in order to support compliance assessment. The crack front was modeled as a semi-ellipse, and the compliance results agreed with experimental data from the literature. It was shown that for the same equivalent physical straight crack standardized by ASTM, compliance generally decreases as tunneling increases. Since the maximum crack curvature allowed by the aforementioned standards is very restrictive, compliance did not meaningfully change within that limit, however, if violated, this paper shows that higher deviations may occur, leading to inaccurate crack depth estimations and invalid test results. These limits and deviations were clearly determined and, as a step to improve the techniques, this paper also presents – in part two – an exploration of a possible approach to mitigate this problem, which is based on the modification of how the equivalent straight crack of a curved crack front is determined. This new approach presents reduced errors in compliance-based crack size estimation as crack curvature increases when compared to current standardized protocols, and it can support further investigations in order to validate and standardize improved measuring techniques. Finally, it is important to state that even though the ASTM E1820 is used for the determination of crack driving forces, this study is based only on the study of the crack front curvature, the limit imposed by this standard and the deviations on crack size estimation when those limits are violated, while not focusing on determining errors directly on the J-integral. This paper is a further development on the studies published before by the research group.
  • Artigo IPEN-doc 29916
    Evaluation of the influence of the viscous sublayer on the mechanical stability of fuel plates under axial flow conditions
    2023 - MOURA, A.J.S.; MATTAR NETO, M.
    The current work aims to investigate the influence of the viscous sublayer on the mechanical stability of fuel element plates under axial flow conditions by means of two-way Fluid Structure Interaction (FSI) numerical simulations. The methodology adopted is that proposed by (Mantec´on, 2019; Mantec´on and Mattar Neto, 2018), who observed a transition from linear to non-linear behavior between the maximum deflection of the plates in their leading edge with the square of the velocity of the cooling fluid in the channel. The speed at which the transition is identified is the critical speed (Vc). In order to verify the influence of the viscous effects, the CFD domain was discretized from its viscous sublayer. As this approach greatly increases the computational cost, where the characteristics of the flow allowed, symmetry boundary conditions were used. In addition to this approach, it was decided to investigate the ability to solve the FSI problem in steady state. The obtained results confirmed that the boundary layer modeling is sufficient to determine the critical velocity. Furthermore, they also suggest that the steady-state approach and the application of symmetry boundary conditions, where possible, can be used in the design of new fuel elements, supporting traditional methods.
  • Artigo IPEN-doc 29904
    Assessment of the von Mises stresses and stress triaxiality in notches using modified tensile specimens
    Complete understanding of the local stress triaxiality and stress concentration is essential to ensuring structural safety of several structures. A combination of mechanical tests with numerical simulations can be used to obtain this information. One way to study stress triaxiality is by modifying the standard tensile test geometry (ASTM E8) with a notch. Based on previous results from the literature, five notches were chosen: 10, 5, 3, 2, and 1 mm. These geometries were tested, and the results were numerically reproduced using the Abaqus/Explicit 2020 software. The models used were a non-linear model with the Gurson-Tvergaard-Needleman damage model to reproduce the failure. The numerical analyses allowed the assessment of the von Mises stress and stress triaxiality near the notch to compare with the standard smooth specimen. Two instants were considered as crack propagation onset; the instant of the maximum von Mises stress in the element at the center of the specimen, where the failure process begins; and the moment of maximum stress in the true stress x true strain curve. For the von Mises stress analysis, the difference between the curves was small. The stress triaxiality is a better variable to visualize the influences of the notch. When the strain is equal to a 0.07 (instant of the maximum force for the standard specimens), for the smaller notches (1 and 2 mm), there is a region where the effective plastic strain is zero. Consequently, the stress triaxiality is larger in this region than in the center. For the crack propagation onset instant, the plastic strain occurs along the whole transversal section. In this instant, the maximum value of stress triaxiality occurs in the center for all specimens. These results demonstrate that the stress triaxiality changes as the strain increases, i.e., varies with time.
  • Artigo IPEN-doc 29555
    Risk-based design of electric power systems for non-conventional nuclear facilities at shutdown modes
    The work presents a methodology for assessing the safety of electrical system designs for non-conventional nuclear facilities in shutdown. The methodology adopts the core damage frequency as the main risk measure to assess the different architectures of power systems in a non-conventional nuclear facility. Among the reasons is the absence of a specific regulatory basis for this type of installation. The adoption of standards for nuclear power plants by non-conventional nuclear facilities does not take into account the functional and operational particularities of these installations, imposing criteria that are often overestimated, which can even lead to an increase in the financial risk for carrying out the projects. Safety probabilistic analyzes become essential tools for the facilities design and licensing. The modeling and quantification of systems failures in charge of ensuring the nuclear safety of non-conventional nuclear facilities are carried out in the CAFTA software environment. In these studies, the analysis of electrical system configurations and their influence on the overall risk of the installation stand out.
  • Artigo IPEN-doc 29552
    External Events PSA
    Since the Fukushima Daiichi accident, external events analysis has become a priority issue within regulatory bodies, operators, and designers, raising concerns about the capabilities of nuclear power plants to withstand severe conditions. Generally, the methodology applied to the Probabilistic Safety Assessment (PSA) of external events consists of the identification of potential single and combined external hazards, screening of external hazards, analysis of site and plant response, analysis of initiating events and quantification of accident sequences probabilities. Therefore, in this paper, the requirements and other information on new nuclear installations projects necessary to implement a comprehensive PSA of external events throughout plant lifetime are evaluated. In addition, it is necessary to clearly identify all the resources that must be available to continuously expand PSA scope to include all types of initiating events, levels of analysis and plant operation modes.
  • Artigo IPEN-doc 29551
    IEA-R1 renewed primary system pump B1-B nozzles stress analysis
    The present report is a summary of the structural analysis of the pump nozzles applying the finite element method by using the Ansys computer program. The IEA-R1 RR is an open pool-type moderated and cooled by light water using beryllium/graphite as a reflector. The reactor can reach up to 5MW of thermal power cooled by the primary and secondary systems. The primary coolant system consists of a piping arrangement, a decay tank, two pumps, and two heat exchangers. The primary pump B1-B presented some failures requiring refurbishment by a new one. The pump used in the IEA-R1 must meet the requirements inherent to the nuclear installation, in addition to the operational requirements for rotating equipment, such as flow and pressure, and structural integrity of the body and nozzles. The supplier specified the type of pump suitable for the System. The pump furnished granted mechanical allowable loads for the nozzles that were lower than the loads imposed by the piping on the nozzles. To enable the installation of the pump in the primary circuit, new support was inserted in the piping system next to the pump minimizing efforts and deformations. A piping stress analysis was carried out to obtain the new efforts imposed on the nozzles. For validation of the motor pump set, a verification of the nozzles was done compared with API 610 standard loads, and the allowable loads of the provider. Finally, a structural analysis of the pump nozzles with the new loads was developed using the finite element method. The calculated stresses meet the limits prescribed by the ASME code; therefore, the new B1-B Pump is approved for operation at the IEA-R1 Nuclear Research Reactor primary circuit.
  • Resumo IPEN-doc 29420
    Assessment of the von Mises stresses and stress triaxiality in notches using modified tensile specimens
    2022 - PEREIRA, L.d.; DONATO, G.H.; MATTAR NETO, M.
    Stress triaxiality is important in fracture mechanics to check the safety of several structures. Stress triaxiality is one of the main factors that influence the fracture process of high toughness steels. For example, a ductile fracture tends to be more predominant for a low constrain geometry with less plastic restriction. The configuration and loading of the structural components are different from those of the mechanical test specimens used to obtain the materials fracture properties. So, understanding the local stress triaxiality is essential to ensure structural safety. Combination of tests with numerical simulations is a way to assess this effect. Modifying the standard tensile test geometry (ASTM E8) with a notch causes a change in the stress triaxiality. Based on the literature information, two notches were chosen: 1 and 2 mm. These geometries were tested, and the results were numerically reproduced using a non-linear model with the GTN damage model in the software Abaqus/Explicit 2020. The properties (elastic and plastic) were obtained from the standard specimen. An axisymmetric finite element model was developed considering the symmetry in the specimen longitudinal direction, and a mesh with the smallest element having the dimensions of 0.2x0.4 mm. First, a test speed of 0.015 mm/s was applied in the specimen longitudinal direction and convergence problems occurred. Thus, the speed was increased to 100 mm/s to solve these problems. Finally, the nine GTN damage parameters were calibrated to describe numerically the experimental curve. The stresses were obtained for the centroid of the elements. All the analyses were done for two points, i.e.,. first is the plastic instability point for a standard specimen, and second is the maximum force of the load vs. displacement curve. The numerical results analysis allowed the assessment of the stress field and stress triaxiality near the notch to compare with the standard specimen. The notch influences the stress locally, but, after a short distance, approximately 45 mm in these specimens, the tendency was the same for three geometries. The evaluation of the triaxiality considered the stress in the specimen longitudinal direction and the hydrostatic stress. Before the point of plastic instability (first point), the stress triaxiality is low, practically an uniaxial stress state. To the second point, the stress state is no longer uniaxial. The notch increases the stress triaxiality across the cross-section, and the biggest value occurred in the center of the specimen. These specimens results can help to identify the region affected by the notches in structural components.
  • Resumo IPEN-doc 29197
    The use of miniaturized samples to determine mechanical properties of materials
    The caracterization of irradiated materials through the SPT (Small Punch Test) technique uses miniaturized samples, with 8 mm in diameter and 0.5 mm in thickness, which has fixed edges, pressed by a sphere that has a diameter d=2.5 mm[1], tested in convencional mechanical testing machines, with the aid of a device developed for their achivement. This tecnique developed for nuclear industry can be used where conventional methods do not apply because it is considered an almost “non-destructive” method[2]due to the small sample volume. In this work two different devices were developed to perform tests at room and sub-zero temperature. The SPT tests will be carried out on standardized nuclear materials unirradiated (ferritic and stainless steels) for later correlation with conventional mechanical tests. Several mechanical properties will be obtained such as yield stress, tensile strength and fracture properties of the materials such as its toughness.