ANTONIO TEIXEIRA E SILVA

Resumo

Graduated in Electronic Engineering from the Federal University of Rio de Janeiro (1975), master's degree in Nuclear Engineering from the Military Institute of Engineering (1978) and doctorate in Nuclear Engineering - Rheinisch-Westfalischen Technischen Hochschule / Aachen (1983) in Germany. He is a full professor at the Energy and Nuclear Research Institute, acting as professor of postgraduate courses since 1985. He has experience in Nuclear Engineering, with an emphasis on Nuclear Fuel and Nuclear Safety Engineering, acting mainly on the following topics: nuclear fuel, nuclear engineering, nuclear fuel design for research and power reactors, nuclear fuel irradiation performance, safety culture, control of nuclear material and safeguards, physical protection and radiation protection. He is the current Safety and Security Coordinator at IPEN / CNEN. (Text obtained from the Currículo Lattes on October 4th 2021)


Possui graduação em Engenharia Eletrônica pela Universidade Federal do Rio de Janeiro (1975), mestrado em Engenharia Nuclear pelo Instituto Militar de Engenharia (1978) e doutorado em Engenharia Nuclear pela Rheinisch-Westfalischen Technischen Hochschule/Aachen (1983) na Alemanha. É professor titular do Instituto de Pesquisas Energéticas e Nucleares, atuando como professor de disciplinas de pós-graduação desde 1985. Tem experiência na área de Engenharia Nuclear, com ênfase na Engenharia do Combustível Nuclear e na Segurança Nuclear e Proteção Física, atuando principalmente nos seguintes temas: combustível nuclear, engenharia nuclear, projeto de combustíveis nucleares para reatores de pesquisa e potência, desempenho sob irradiação do combustível nuclear, cultura de segurança, controle de material nuclear e salvaguardas, proteção física e proteção radiológica. É o atual Coordenador de Segurança do IPEN-CNEN. (Texto extraído do Currículo Lattes em 04 out. 2021)

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Resultados de Busca

Agora exibindo 1 - 10 de 22
  • Artigo IPEN-doc 29914
    Assessment of minimum allowable thickness of advanced steel (FeCrAl) cladding for accident tolerant fuel
    2023 - ABE, ALFREDO; GIOVEDI, CLAUDIA; MELO, CAIO; SILVA, ANTONIO T. e
    The ferritic iron-chromium-aluminum (FeCrAl) alloy cladding is considered to be the most promising for near-term application in the ATF framework to replace existing zirconium alloy cladding. Although FeCrAl cladding presents several advantages, it is well known that there are at least two main drawbacks, one is the increased thermal neutron absorption cross-section compared to the current Zr-based cladding resulting in a neutronic penalty and another is tritium higher permeation. In the present study, the minimum allowable thickness of cladding is addressed considering neutronic penalty reduction and the mechanical-structural behavior under the LOCA accident condition. The neutronic penalty assessment was performed using the Monte Carlo code and mechanical-structural performance of the FeCrAl cladding using the TRANSURANUS fuel code, which was modified to consider properly the FeCrAl cladding.
  • Artigo IPEN-doc 28507
    A method for uncertainty and sensitivity analysis in fuel performance codes
    2021 - DANTAS, A.C.; SILVA, A.T.
    The present study proposes a method for the execution of uncertainty and sensitivity analysis on TRANSURANUS code, adapted for the use of stainless steel AISI-348 as the cladding material for a PWR reactor fuel rod, thus allowing to determine which input data are more relevant to the TRANSURANUS models, as well as a confidence interval for the results. The analysis was made through Monte Carlo sampling, where input values related to the geometry and composition of the fuel rod were taken from a normal distribution truncated around fabrication tolerance values. The generated samples were used as TRANSURANUS input data, and after numerous executions of the code, the results pertaining to the fuel center line temperature, fuel rod inner pressure and cladding strains were used to obtain a confidence interval and to make a variance-based sensitivity analysis, showing that the models used in TRANSURANUS are additive in nature, and input interactions are not relevant to the code.
  • Artigo IPEN-doc 27620
    The IPEN/CNEN contribution to IAEA FUMAC benchmark using modified fuel performance code based on stainless steel as cladding under steady state, transient and accident conditions
    2020 - ABE, ALFREDO; SILVA, ANTONIO T. e; GIOVEDI, CLAUDIA; MELO, CAIO; GOMES, DANIEL de S.; MUNIZ, RAFAEL R.
    The IPEN/CNEN (Brazil) participated in IAEA Coordinated Research Project on Fuel Modeling in Accident Conditions (FUMAC) among others 18 countries (Argentina, Belgium, Bulgaria, China, Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Norway, Republic of Korea , Russian Federation , Spain , Sweden , Ukraine and United States of America), which aim was focused in modelling, predicting and improving the understanding of the behaviour of nuclear fuel under accident conditions in order to better understanding and enhanced safety of nuclear fuel. A serie of LOCA (Loss of Coolant Accident) experiments data were made available for the participants to perform simulation using their fuel performance codes and the outcome gives an idea about fuel codes limitation considering LOCA simulation and possible improvement needed in the existing models related to LOCA condition.The IPEN/CNEN (BRAZIL) proposal for FUMAC-CRP was to modify existing fuel performance codes (FRAPCON and FRAPTRAN) considering stainless steel as cladding material and perform a simulation comparing to zircaloy cladding performance under steady state and accident condition. The HALDEN LOCA Experiments (IFA 650-9, IFA-650-10 and IFA-650-11) were selected and modeled to perform the LOCA accident simulation considering the original cladding (zircaloy) and compared to stainless steel cladding.
  • Artigo IPEN-doc 26855
    Reactivity initiated accident assessment for ATF cladding materials
    2020 - GIOVEDI, C.; MARTINS, M.R.; ABE, A.; REIS, R.; SILVA, A.T.
    Following the experience that came from the Fukushima Daiichi accident, one possible way of reducing risk in a nuclear power plant operation would be the replacement of the existing fuel rod cladding material (based on zirconium alloys) by another materials which could fulfill the requirements of the accident tolerant fuel (ATF) concept. In this sense, ATF should be able to keep the current fuel system performance under normal operation conditions; moreover, it should present superior performance than the existing conventional fuel system (zirconium-based alloys and uranium dioxide) under accident conditions. The most challenging and bounding accident scenarios for nuclear fuel systems in Pressurized Water Reactors (PWR) are Loss of Coolant Accident (LOCA) and Reactivity Initiated Accident (RIA), which are postulated accidents. This work addresses the performance of ATF using iron-based alloys as cladding material under RIA conditions. The evaluation is carried out using modified versions of the coupled system FRAPCON/FRAPTRAN. These codes were modified to include the material properties (thermal, mechanical, and physics) of an iron-based alloy, specifically FeCrAl alloy. The analysis is performed using data available in the open literature related to experiments using conventional PWR fuel system (zirconium-based alloys and uranium dioxide). The results obtained using the modified code versions are compared to those of the actual existing fuel system based on zircaloy-4 cladding using the original versions of the fuel performance codes (FRAPCON/FRAPTRAN).
  • Artigo IPEN-doc 26854
    Assessment of high conductivity ceramic fuel concept under normal and accident conditions
    2020 - GOMES, D.S.; ABE, A.; SILVA, A.T.; MUNIZ, R.O.R.; GIOVEDI, C.; MARTINS, M.R.
    After the Fukushima Daiichi accident, the high conductivity ceramic concept fuel has been revisited. The thermal conductivity of uranium dioxide used as nuclear fuel is relatively low, as consequence fuel pellet centerline reaches high temperatures, high fission gas release rate, increase of fuel rod internal pressure reducing the safety thermal margin. Several investigations had been conducted in framework of ATF (Accident Tolerant Fuel) using different additives in ceramic fuel (UO2) in order to enhance thermal conductivity in uranium dioxide pellets. The increase of the thermal conductivity of fuel can reduce the pellet centerline temperature, consequently less fission gas releasing rate and the low risk of fuel melting, hence improving significantly fuel performance under accident conditions. The beryllium oxide (BeO) has high conductivity among other ceramics and is quite compatible with UO2up to 2200°C, at which temperature it forms a eutectic. Moreover, it is compatible with zircaloy cladding, does not react with water, has a good neutronic characteristics (low neutron absorption cross-section, neutron moderation). This work presents a preliminary assessment of high conductivity ceramic concept fuel considering UO2-BeO mixed oxide fuel containing 10 wt% of BeO. The FRAPCON and FRAPTRAN fuel performance codes were conveniently adapted to support the evaluation of UO2-BeO mixed oxide fuel. The thermal and mechanical properties were modified in the codes for a proper and representative simulation of the fuel performance. Theobtainedpreliminary results show lower fuel centerline temperatureswhen compared to standard UO2 fuel, consequently promoting enhancement of safety margins during the operational condition and under LOCA accident scenario.
  • Capítulo IPEN-doc 26711
    Development and application of modified fuel performance code based on stainless steel as cladding under steady state, transient and accident conditions
    2019 - ABE, ALFREDO; SILVA, ANTONIO T. e; GIOVEDI, CLAUDIA; MELO, CAIO; GOMES, DANIEL de S.; MUNIZ, RAFAEL R.
    The IPEN/CNEN proposal for FUMAC-CRP was to modified fuel performance codes (FRAPCON and FRAPTRAN) in order to assess the behavior of fuel rod using stainless steel as cladding and compare to zircaloy cladding performance under steady state and accident condition. The IFA 650- 9, IFA-650-10 and UFA-650-11experiments were modelled to perform the LOCA accident simulation considering the original cladding and compared to stainless steel cladding.
  • Artigo IPEN-doc 26363
    Modification of TRANSURANUS fuel performance code in the ATF framework
    2019 - ABE, ALFREDO Y.; MELO, CAIO; GIOVEDI, CLAUDIA; SILVA, ANTONIO T.
    The standard fuel system based on UO2–zirconium alloy has been utilized on nearly 90% of worldwide nuclear power light water reactors. After the Fukushima Daiichi accident, alternative cladding materials to zirconium-based alloys are being investigated in the framework of accident tolerance fuel (ATF) program. One of the concepts of ATF is related to cladding materials that could delay the onset of high temperature oxidation, as well as ballooning and burst, in order to improve reactor safety systems, and consequently increase the coping time for the reactor operators in accident condition, especially under Loss-of-Coolant Accident (LOCA) scenario. The ferritic iron-chromium-aluminum (FeCrAl) alloys have been identified as an alternative to replace current zirconium-based alloys based on its outstanding resistance to oxidation under superheated steam environment due to the development of alumina oxide on the alloy surface in case of LOCA; moreover, FeCrAl alloys present quite well performance under normal operation conditions due to the thin oxide rich in chromium that acts as a protective layer. The assessment and performance of new fuel systems rely on experimental irradiation program and fuel performance code simulation, therefore the aim of this work is to contribute to the computational modeling capabilities in the framework of the ATF concept. The well-known TRANSURANUS fuel performance code that is used by safety authorities, industries, laboratories, research centers and universities was modified in order to support FeCrAl alloy as cladding material. The modification of the TRANSURANUS code was based on existing data (material properties) from open literature and as verification process was performed considering LOCA accident scenario.
  • Artigo IPEN-doc 26356
    Fuel performance of iron-based alloy cladding using modified TRANSURANUS code
    2019 - GIOVEDI, CLAUDIA; MELO, CAIO; ABE, ALFREDO Y.; SILVA, ANTONIO T.; MARTINS, MARCELO R.
    The main challenge in the nuclear area since the Fukushima Daiichi accident is to develop fuel materials to be applied in nuclear reactors aiming to increase the safety under normal operation as well as transient and accident conditions. These efforts are concentrated in the Advanced Technology Fuel (ATF) program that has as main scopes to study cladding materials to replace the zirconium-based alloys, and fuel materials presenting higher thermal conductivity compared to the conventional uranium dioxide fuel pellet. In this sense, iron-based alloys, which were used with a good performance as cladding material in the first Pressurized Water Reactors (PWR), have becoming a good option. The assessment of the behavior of different materials previously to perform irradiation tests, which are time consuming, can be performed using fuel performance codes, but for this, the conventional fuel performance codes must be modified to implement the properties of the materials that are being studied. This paper presents the results obtained using a modified version of the well-known TRANSURANUS code, obtained from the implementation of the stainless steel 348 properties as cladding material. The simulations were performed using data available in the open literature related to a PWR irradiation experiment. The results obtained using the modified version of the code were compared to those obtained using the original code version for zircaloy-4. The performance of both cladding materials was evaluated by means of the comparison of parameters such as gap thickness, fuel centerline temperature, internal pressure, and cladding stress and strain.
  • Artigo IPEN-doc 26269
    Mechanical-structural analysis of a stainless steel fuel rod under burst test conditions
    2019 - FARIA, DANILO P.; SILVA, ANTONIO T. e; LIMA, LEONARDO S.; BERRETTA, JOSE R.
    After the Fukushima nuclear accident in 2011, the nuclear research community has initiated research into the development of fuels that are resistant to accidents. In this context, iron-based alloys have emerged as a good alternative to zirconium alloys. In order to make possible the cladding material replacement, studies related to their mechanical properties are necessary. Thus, the present study carried out a mechanical-structural evaluation from the available data collection regarding the mechanical properties of stainless steel 348, specifically in the conditions of the burst test. Burst tests were performed at various temperatures ranging from 32°C up to 450 °C. Then, a computational model was created based on the specimen of the burst test. Numerical simulation was performed considering the tensile tests of stainless steel at various temperatures. The numerical results were compared with the results of the burst test. Test and simulations were comparable leading to computational model validation. As austenitic stainless steels have structural stability for low and high temperatures, the results could be extrapolated to temperatures higher than those in the burst test. After the validation of the computational model, simulations were performed for temperatures higher than 450ºC, thus obtaining a burst pressure curve as a function of the temperature for stainless steel ANSI 348. The correlation of burst data as function of temperature could be implemented in the FRAPTRAN code, in order to make possible the evaluation of the behavior of a fuel rod with stainless steel ANSI 348 under postulated accident conditions (LOCA).