ALFREDO YUUITIRO ABE
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Tese IPEN-doc 29731 Análise neutrônica e do comportamento sob irradiação de combustíveis tolerantes à falha2023 - ABE, ALFREDO Y.O presente trabalho apresenta uma avaliação e análise quanto aos aspectos neutrônicos e de desempenho de combustível para os potenciais candidatos a combustíveis tolerantes a falha (ATF). Na fase inicial do trabalho para o levantamento bibliográfico utilizou-se a metodologia baseada na técnica de maturidade tecnológica (TRL) para avaliar as diversas opções e tecnologias desses combustíveis. Esta fase inicial permitiu identificar nos desenvolvimentos dos combustíveis tolerantes a falha as principais lacunas e as atividades necessárias para a obtenção e utilização comercial em reatores de potência. As avaliações e as análises neutrônica e de desempenho do combustível dos principais candidatos à combustíveis tolerantes a falha foram realizadas por meio de simulações computacionais utilizando programas de desempenho do combustível (TRANSURANUS) e neutrônica (SERPENT). As atividades consistiram na avaliação e verificação de um projeto de núcleo de reator constituído de combustíveis tolerantes a falha (combustíveis: UO2, U3Si2, UN, UO2-BeO) e os revestimentos: ZIRLO, SiC, FeCrAl, AISI-348). Para tanto, foram obtidos e avaliados os principais parâmetros associados ao projeto do núcleo do reator. Além disso, o trabalho envolveu a avaliação de desempenho de um dos revestimentos (FeCrAl) do combustível tolerantes a falha mais promissores considerando as condições de operação normal e em cenário de acidente. Como conclusão preliminar das atividades desenvolvidas neste trabalho destaca-se a viabilidade neutrônica dos diferentes combustíveis ATF, no entanto em termos de desempenho do combustível nem todas as opções disponíveis ainda possuem grau de maturidade suficiente para aplicação na indústria nuclear em curto e médio prazo. Destaca-se que este trabalho aborda e busca contribuir com um tema bastante atual e relevante na comunidade nuclear, principalmente os grandes desafios envolvidos no desenvolvimento, fabricação e o licenciamento dos combustíveis tolerantes a falha junto aos órgãos regulatórios.Artigo IPEN-doc 28287 The FeCrAl cladding assessment under accident condition using TRANSURANUS fuel performance code2021 - ABE, ALFREDO; MELLO, CAIO; SANTOS, TAMIRYS; GIOVEDI, CLAUDIAArtigo IPEN-doc 28256 Passive Autocatalytic Recombiner perfomance assessment using COCOSYS code2021 - GALVAO, H.P.; SHORTO, J.M.B.; SOBRINHO, G.T.; ABE, A.Y.; GIOVEDI, CArtigo IPEN-doc 27926 Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario2021 - 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.Artigo IPEN-doc 26356 Fuel performance of iron-based alloy cladding using modified TRANSURANUS code2019 - 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 26355 Sensitivity analysis of fuel rod parameters in steady state condition using TRANSURANUS code2019 - AGUIAR, AMANDA A.; ABE, ALFREDO; GIOVEDI, CLAUDIAIn this paper, a simulation of steady state conditions using TRANSURANUS code applied to Arkansas Nuclear One Unit 2 (PWR) fuel rod is presented. The fuel rod considered in this work was exposed to a peak rod average burnup of 64 GWd/TU, which corresponds to a batch-average exposure of about 53 GWd/TU. TRANSURANUS code offers two different approach for sensitivity analysis: Numerical Noise Analysis and Monte Carlo. In this work, sensitivity analysis using Monte Carlo approach was considered in the range of fuel rod manufacturing parameters, such as internal and external radius of the cladding, external radius of the fuel, and filling gas pressure of the fuel rod, in order to verify some existing correlation with fuel centerline temperature, internal cladding temperature, average tangential stress in the cladding, average permanent tangential strain in the cladding, internal pressure, and fission gas release.Artigo IPEN-doc 24021 Sensitivity assessment of fuel performance codes for loca accident scenario2017 - ABE, ALFREDO; GIOVEDI, CLAUDIA; GOMES, DANIEL; SILVA, ANTONIO T. e; MUNIZ, RAFAEL O.R.; MARTINS, MARCELOFRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment.Artigo IPEN-doc 24013 Simulation of accident-tolerant U3Si2 fuel using FRAPCON code2017 - GOMES, DANIEL S.; SILVA, ANTONIO T.; ABE, ALFREDO Y.; MUNIZ, RAFAEL O.R.; GIOVEDI, CLAUDIAThe research on accident-tolerant fuels (ATFs) increased after the Fukushima event. This benefitted risk management in nuclear operations. In this investigation, the physical properties of the materials being developed for the ATF program were compared with those of the standard UO2–Zr fuel system. The research efforts in innovative fuel design include rigorous characterization of thermal, mechanical, and chemical assessment, with the objectives of making the burnup cycle longer, increasing power density, and improving safety performance. Fuels must reach a high uranium density—above that supported by UO2—and possess coating that exhibits better oxidation resistance than Zircaloy. The uranium density and thermal conductivity of ATFs, such as U3Si2, UN, and UC, is higher than that of UO2; their combination with advanced cladding provides possible fuel–cladding options. An ideal combination of fuel and cladding must increase fuel performance in loss-of-coolant scenarios. The disadvantages of U3Si2, UN, and UC are their swelling rates, which are higher than that of UO2. The thermal conductivities of ATFs are approximately four times higher than that of UO2. To prevent the generation of hydrogen due to oxidation of zirconium-based alloys in contact with steam, cladding options, such as ferritic alloys, were studied. It was verified that FeCrAl alloys and SiC provide better response under severe conditions because of their thermophysical properties. The findings of this study indicate that U3Si2 and the FeCrAl fuel cladding concept should replace UO2–Zr as the fuel system of choice.Artigo IPEN-doc 24012 Modification of fuel performance code to evaluate iron-based alloy behavior under loca scenario2017 - GIOVEDI, CLAUDIA; ABE, ALFREDO; MUNIZ, RAFAEL O.R.; GOMES, DANIEL de S.; SILVA, ANTONIO T. e; MARTINS, MARCELO 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 a fuel performance code to evaluate the behavior of ironbased 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.Artigo IPEN-doc 24011 Assessment of uranium dioxide fuel performance with the addition of beryllium oxide2017 - MUNIZ, RAFAEL O.R.; GIOVEDI, CLAUDIA; ABE, ALFREDO; GOMES, DANIEL S.; AGUIAR, AMANDA A.; SILVA, ANTONIO T.The Fukushima Daiichi accident in 2011 pointed the problem related to the hydrogen generation under accident scenarios due to the oxidation of zirconium-based alloys widely used as fuel rod cladding in water-cooled reactors. This problem promoted research programs aiming the development of accident tolerant fuels (ATF) which are fuels that under accident conditions could keep longer its integrity enabling the mitigation of the accident effects. In the framework of the ATF program, different materials have been studied to be applied as cladding to replace zirconium-based alloy; also efforts have been made to improve the uranium dioxide thermal conductivity doping the fuel pellet. This paper evaluates the addition of beryllium oxide (BeO) to the uranium dioxide in order to enhance the thermal conductivity of the fuel pellet. Investigations performed in this area considering the addition of 10% in volume of BeO, resulting in the UO2-BeO fuel, have shown good results with the improvement of the fuel thermal conductivity and the consequent reduction of the fuel temperatures under irradiation. In this paper, two models obtained from open literature for the thermal conductivity of UO2- BeO fuel were implemented in the FRAPCON 3.5 code and the results obtained using the modified code versions were compared. The simulations were carried out using a case available in the code documentation related to a typical pressurized water reactor (PWR) fuel rod irradiated under steady state condition. The results show that the fuel centerline temperatures decrease with the addition of BeO, when compared to the conventional UO2 pellet, independent of the model applied.