ANTONIO TEIXEIRA E SILVA

ANTONIO TEIXEIRA E SILVA

(Fonte: Lattes)

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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Potential Organizational Behavior Management (OBM) contributions for raising Computer Security awareness and insider threat mitigation
2023 - BIANCHI, PAULO H.; PESSOA, CANDIDO V.B.B.; SILVA, ANTONIO T. e
Organizational Behavior Management (OBM) is a research field dedicated for developing processes to modify human behavior in organizational environment. It is derived from Behavior Analysis, a methodology for studying human behavior with three characteristics that enables research to be translated into applied technology: Quantification, variables can be quantified and standardized; Repetition, results are predictable in a degree of trustworthiness; and Verification, processes are described with sufficient details allowing its replication. For OBM, an organization applies Value Based Governance when it enables and reinforces employees to change their environment according to organizational values. Also, for OBM, Verbal Governance is the leadership’s capacity to verbally engage and motivate employees to comply with organizational rules and values. The IAEA NSS No. 42-G states that a security culture is an essential aspect of any nuclear security regime and Computer Security should be accounted when promoting security culture in nuclear facilities. In this work we will argue that Value Based Governance would mitigate insider threat, especially from disgruntled employees, and would also raise awareness when Computer Security is an important organizational value for leadership. Finally, we suggest an effective, evidence-based Verbal Governance technique able to promote Computer Security values in nuclear facilities.
Strategies for decommissioning small nuclear reactors in Brazil
2023 - CALDAS NETO, A.B.; SILVA, A.T.
The process of decommissioning nuclear reactors is a complex activity that involves various technical and administrative stages. Its main objective is to ensure the safety of the site, workers, the general public, and the environment during the execution of decommissioning activities, aiming for the release of the site from regulatory control. In the Brazilian context, it is essential to develop decommissioning strategies, taking into consideration the established technical and regulatory requirements, as well as following the guidelines of the Brazilian Nuclear Policy (BNP). Eight decommissioning strategies were proposed for small reactors, with different objectives and in different scenarios, encompassing 23 decommissioning approaches, divided into 6 groups: 1) decontamination and dismantling (DD); 2) radioactive waste (RW) management; 3) RW storage management; 4) human resources (HR) and knowledge management; 5) cost estimation; and 6) financial fund management. Additionally, 18 factors affecting the selection of these approaches were considered, taking into account particularities of the Brazilian context. A qualitative risk analysis was conducted using risk assessment techniques from the ABNT NBR ISO/IEC 31,010 standard, with a focus on the Multicriteria Decision Analysis (MCDA) technique. This qualitative analysis allowed for the evaluation of the approaches considering the current scenario and the future scenario, which includes possible changes in the BNP currently under discussion in the National Congress. The observations and results obtained in this study will be useful in guiding future efforts related to nuclear reactor decommissioning projects in Brazil. Based on the proposed strategies and considerations of safety, regulation, and governmental policies, it will be possible to plan and execute decommissioning activities more efficiently and safely.
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.
MEA - Modified Energy Amplifier proposal
2001 - MAIORINO, J.R.; PEREIRA, S.A.; SANTOS, A.; SILVA, A.T.
Recently Rubbia et al proposed a conceptual design of an Accelerator Driven System, known as Energy Amplifier (EA), as an advanced innovative reactor which utilizes a spallation neutron source induced by protons, from a Cyclotron or Linac, in a subcritical array imbibed in liquid lead coolant. Besides of being breeder and waste burner, the conceptual design generates energy and allows the use of Thorium as fuel. This paper introduces some qualitative changes in the Rubbia's concept. More than one point of spallation is proposed in order to reduce the requirement of proton energy and current of the accelerator, and mainly to make a flatter power density distribution. The subcritical core, which in the Rubbia's concept is an hexagonal array of pins immersed in a liquid lead coolant, is replaced by a concept of a solid lead calandria with the fuel elements in channels cooled by Helium. This concept allows on line refueling or shuffling, and the utilization of a direct thermodynamic cycle (Brayton), which is more efficient than a vapor cycle. Although the calculations to demonstrate the feasibility of the MEA concept are underway and not yet finished, these ideas do not violate the basic physics of the EA, but reduces requirement in the accelerator complex, which is more realistic and economical in today accelerators technology. Finally, the utilization of He as coolant, compared with liquid Pb, is more realistic since the gas cooled reactors technology is well established and more efficient from the thermodynamic view, allowing simplification and the utilization in high temperature process, like hydrogen generation.
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.
Computer security on Brazilian nuclear facilities
2021 - TAVARES, R.L.A.; LEMOS, F.L.; SILVA, A.T.
Development of a computer code PADPLAC-UMo for performance analysis of monolithic uranium molybdenum fuel plate
2021 - SILVA, A.T. e; REIS, R.
Preliminary assessment of iron alloy cladding as accident tolerant fuel cladding
2019 - ABE, ALFREDO; TEIXEIRA, ANTONIO; SOUZA, DANIEL; GIOVEDI, CLAUDIA
Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario
2021 - GIOVEDI, C.; ABE, A.; MUNIZ, R.O.R.; GOMES, D.S.; SILVA, A.T.; MARTINS, M.R.
Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in FRAPCON and FRAPTRAN fuel performance codes to evaluate the behavior of iron-based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes.
Comparison of computer programs to analyze the irradiation performance of U-Mo monolithic fuel plates and UO2 cylindrical fuel rods in power reactors
2021 - SILVA, A.L.C.; SILVA, A.T.
The aim of this work is to present a comparative analysis in terms of the irradiation performance of cylindrical uranium dioxide fuel rods and monolithic uranium molybdenum fuel plates in pressurized light water reactors. To analyze the irradiation performance of monolithic uranium molybdenum fuel plates when subjected to steady state operating conditions in light water pressurized reactors, the computer program PADPLAC-UMo was used, which performs thermal and mechanical analysis of the fuel taking into account the physical , chemicals and irradiation effects to which this fuel is subjected. For the analysis of the uranium dioxide fuel rods, the code FRAPCON was used, which is an analytical tool that verifies the irradiation performance of fuel rods of pressurized light water reactor, when the power variations and the boundary conditions are slow enough for the term permanent regime to be applied. The analysis for a small nuclear power reactor, despite the higher power density applied to the fuel plate in relation to the fuel rod, showed that the fuel plates have lower temperatures and lower fission gas releases throughout the analyzed power history, allowing the use of a more compact reactor core without exceeding the design limits imposed on nuclear fuel.