Navegação Eventos - Artigos por autor "MUNIZ, RAFAEL O.R."

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  • IPEN-DOC 24947

    GIOVEDI, CLAUDIA; ABE, ALFREDO ; MUNIZ, RAFAEL O.R. ; GOMES, DANIEL S. ; SILVA, ANTONIO T. e ; MARTINS, MARCELO R.. Analysis of the combined effects on the fuel performance of UO2-BeO as fuel and iron-based alloy as cladding. In: WATER REACTOR FUEL PERFORMANCE MEETING, September 10-14, 2017, Jeju Island, Korea. Proceedings... 2017. p. 1-9.

    Abstract: Iron-based alloys have been considered as promising candidate material to replace zirconium-based alloys as fuel cladding based on the previous experience of the first generation of pressurized water reactors (PWR). Moreover, the safety margins of nuclear fuels can be improved by means of additives in the fuel pellet, as beryllium oxide (BeO), due to the increase of the fuel thermal conductivity. These efforts are part of the accident tolerant fuel (ATF) program which aims to develop nuclear fuel systems with enhanced performance under normal operation, design-basis accident and severe-accident conditions. This paper addresses the combined effects on the fuel performance considering the BeO additive in the fuel pellet and stainless steel 348 as cladding material under steady-state and loss-of-coolant-accident (LOCA) scenario. The fuel performance simulation and assessment are conducted using modified versions of well-known fuel performance codes (FRAPCON/FRAPTRAN). The obtained results have shown that the studied fuel system (stainless steel cladding and UO2-BeO) enables an improvement in the main parameters associated to the fuel safety margins under steady-state irradiation as well as LOCA scenario.

  • IPEN-DOC 24009

    GOMES, DANIEL S. ; ABE, ALFREDO Y. ; MUNIZ, RAFAEL O.R. ; GIOVEDI, CLAUDIA. Analysis of UO2-BEO fuel under transient using fuel performance code. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 22-27, 2017, Belo Horizonte, MG. Proceedings... Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear, 2017.

    Abstract: Recent research has appointed the need to replace the classic fuel concept, used in light water reactors. Uranium dioxide has a weak point due to the low thermal conductivity, that produce high temperatures on the fuel. The ceramic composite fuel formed of uranium dioxide (UO2), with the addition of beryllium oxide (BeO), presents high thermal conductivity compared with UO2. The oxidation of zirconium generates hydrogen gas that can create a detonation condition. One of the preferred options are the ferritic alloys formed of iron-chromium and aluminum (FeCrAl), that should avoid the hydrogen release due to oxidation. In general, the FeCrAl alloys containing 10–20Cr, 3–5Al, and 0–0.12Y in weight percent. The FeCrAl alloys should exhibit a slow oxidation kinetics due to chemical composition. Resistance to oxidation in the presence of steam is improved as a function of the content of chromium and aluminum. In this way, the thermal and mechanical properties of the UO2-BeO-10%vol, composite fuel were coupled with FeCrAl alloys and added to the fuel codes. In this work, we examine the fuel rod behavior of UO2-10%vol-BeO/FeCrAl, including a simulated transient of reactivity. The fuels behavior shown reduced temperature with UO2-BeO/Zr, UO2-BeO/FeCrAl also were compared with UO2/Zr system. The case reactivity initiated accident analyzed, reproducing the fuel rod called VA-1 using UO2/Zr alloys and compared with UO2-BeO/FeCrAl.

  • IPEN-DOC 24011

    MUNIZ, RAFAEL O.R. ; GIOVEDI, CLAUDIA; ABE, ALFREDO ; GOMES, DANIEL S. ; AGUIAR, AMANDA A.; SILVA, ANTONIO T. . Assessment of uranium dioxide fuel performance with the addition of beryllium oxide. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 22-27, 2017, Belo Horizonte, MG. Proceedings... Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear, 2017.

    Abstract: 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.

  • IPEN-DOC 15253

    BENITES, DANIELA B.; MUNIZ, RAFAEL O.R.; COELHO, PAULO R.P. . Caracterizacao de dose em campo misto de radiacao utilizando dosimetros termoluminescentes na instalacao para estudos em BNCT. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE; MEETING ON NUCLEAR APPLICATIONS, 9th; MEETING ON REACTOR PHYSICS AND THERMAL HYDRAULICS, 16th; MEETING ON NUCLEAR INDUSTRY, 1st, September 27 - October 2, 2009, Rio de Janeiro, RJ. Proceedings... Sao Paulo: ABEN, 2009, 2009.

    Palavras-Chave: dosimetry; gamma detection; interactions; neutron capture therapy; personnel dosimetry; radiation doses; radiations; radioactivation; thermoluminescence; thermoluminescent dosemeters; thermoluminescent dosimetry

  • IPEN-DOC 15019

    MUNIZ, RAFAEL O.R.; COELHO, PAULO R.P. ; ANDRADE e SILVA, GRACIETE S. . Experimental results analysis amd simulation to evaluate flux and dose at the irradiation sample position of the BNCT research facility. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE; MEETING ON NUCLEAR APPLICATIONS, 9th; MEETING ON REACTOR PHYSICS AND THERMAL HYDRAULICS, 16th; MEETING ON NUCLEAR INDUSTRY, 1st, September 27 - October 2, 2009, Rio de Janeiro, RJ. Proceedings... Sao Paulo: ABEN, 2009, 2009.

    Palavras-Chave: computerized simulation; dose rates; epithermal neutrons; experimental data; gamma radiation; iear-1 reactor; neutron capture therapy; neutron flux; thermal neutrons; thermoluminescent dosemeters

  • IPEN-DOC 24014

    GOMES, DANIEL S. ; SILVA, ANTONIO T. ; ABE, ALFREDO Y. ; MUNIZ, RAFAEL O.R. ; GIOVEDI, CLAUDIA. High density fuels using dispersion and monolithic fuel. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 22-27, 2017, Belo Horizonte, MG. Proceedings... Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear, 2017.

    Abstract: Fuel plates used in high-performance research reactors need to be converted to low-enrichment uranium fuel; the fuel option based on a monolithic formulation requires alloys to contain 6 – 10 wt% Mo. In this case, the fuel plates are composed of the metallic alloy U-10Mo surrounded by a thin zirconium layer encapsulated in aluminum cladding. This study reviewed the physical properties of monolithic forms. The constraints produced during the manufacturing process were analyzed and compared to those of dispersed fuel. The bonding process used for dispersion fuels differs from the techniques applied to foil bonding used for pure alloys. The quality of monolithic plates depends on the fabrication method, which usually involves hot isostatic pressing and the thermal annealing effect of residual stress, which degrades the uranium cubic phase. The preservation of the metastable phase has considerable influence on fuel performance. The physical properties of the foil fuel under irradiation are superior to those of aluminum-dispersed fuels. The fuel meat, using zirconium as the diffusion barrier, prevents the interaction layer from becoming excessively thick. The problem with dispersed fuel is breakaway swelling with a medium fission rate. It has been observed that the fuel dispersed in aluminum was minimized in monolithic forms. The pure alloys exhibited a suitable response from a rate at least twice as much as the fission rate of dispersions. The foils can support fissile material concentration combined with a reduced swelling rate.

  • IPEN-DOC 15280

    SOUZA, GREGORIO S.; MUNIZ, RAFAEL O.R.; COELHO, PAULO R.P. . Improvements at the biological shielding of BNCT research facility in the IEA-R1 reactor. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE; MEETING ON NUCLEAR APPLICATIONS, 9th; MEETING ON REACTOR PHYSICS AND THERMAL HYDRAULICS, 16th; MEETING ON NUCLEAR INDUSTRY, 1st, September 27 - October 2, 2009, Rio de Janeiro, RJ. Proceedings... Sao Paulo: ABEN, 2009, 2009.

    Palavras-Chave: biological shielding; experimental data; iear-1 reactor; lead; neutron capture therapy; radiation doses; radiation protection; thermal neutrons

  • IPEN-DOC 24015

    GOMES, DANIEL S. ; MUNIZ, RAFAEL O.R. ; GIOVEDI, CLAUDIA. Improving performance with accident tolerant-fuels. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 22-27, 2017, Belo Horizonte, MG. Proceedings... Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear, 2017.

    Abstract: After the Fukushima reactor accident, efforts to improve risk management in nuclear operations have included the intensification of research on accident-tolerant fuels (ATFs). In this investigation, the physical properties of recently developed ATFs were compared with those of the current standard fuel, UO2–Zr. The goals for innovative fuel design include a rigorous characterization of the thermal, mechanical, and chemical considerations. The intentions are to lengthen the burnup cycle, raise the power density, and improve safety. Fuels must have a high uranium density—above that supported by UO2—and possess a coating that exhibits better oxidation resistance than Zircaloys. ATFs such as U3Si2, UN, and UC contain a higher uranium density and thermal conductivity than UO2, providing significant benefits. The ideal combination of fuel and cladding must increase performance in a loss-of-coolant accident. However, U3Si2, UN, and UC have a disadvantage; their respective swelling rates are higher than that of UO2. These ATFs also have thermal conductivities approximately four times higher than that of UO2. A study was conducted investigating the hydrogen generated by the oxidation of zirconium alloys in contact with steam using cladding options such as Fe-Cr-Al and silicon carbide. It was confirmed that ferritic alloys offer a better response under severe conditions, because of their mechanical properties as creep rate. The findings of this study indicate that advanced fuels should replace UO2–Zr as the fuel system of choice.

  • IPEN-DOC 16915

    DOMINGOS, DOUGLAS B.; TEIXEIRA e SILVA, ANTONIO ; JOAO, THIAGO G.; MUNIZ, RAFAEL O.R.; COELHO, TALITA S.. Low enriched uranium foil targets with different geometries for the production of molybdenum-99 in the RMB. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE; MEETING ON NUCLEAR APPLICATIONS, 10th; MEETING ON REACTOR PHYSICS AND THERMAL HYDRAULICS, 17th; MEETING ON NUCLEAR INDUSTRY, 2nd, October 24-28, 2011, Belo Horizonte, MG. Proceedings... Sao Paulo: ABEN, 2011, 2011.

    Palavras-Chave: brazilian cnen; ipen-mb-1 reactor; isotope production reactors; targets; molybdenum 99; thermal hydraulics; computer codes

  • IPEN-DOC 24012

    GIOVEDI, CLAUDIA; ABE, ALFREDO ; MUNIZ, RAFAEL O.R. ; GOMES, DANIEL de S. ; SILVA, ANTONIO T. e ; MARTINS, MARCELO R.. Modification of fuel performance code to evaluate iron-based alloy behavior under loca scenario. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 22-27, 2017, Belo Horizonte, MG. Proceedings... Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear, 2017.

    Abstract: 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.

  • IPEN-DOC 19502

    MUNIZ, RAFAEL O.R.; DOMINGOS, DOUGLAS B.; SANTOS, ADIMIR dos ; TEIXEIRA e SILVA, ANTONIO ; JOAO, THIAGO G.; AREDES, VITOR O.. Neutronic comparison of the nuclear fuels Usub(3)Sisub(2)/Al and U-Mo/Al. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE; MEETING ON NUCLEAR APPLICATIONS, 11th; MEETING ON REACTOR PHYSICS AND THERMAL HYDRAULICS, 18th; MEETING ON NUCLEAR INDUSTRY, 3rd, November 24-29, 2013, Recife, PE. Proceedings... Sao Paulo: ABEN, 2013, 2013.

    Palavras-Chave: aluminium alloys; density; dispersion nuclear fuels; fabrication; neutron absorbers; neutron flux; uranium; uranium silicides; uranium-molybdenum fuels

  • IPEN-DOC 24021

    ABE, ALFREDO ; GIOVEDI, CLAUDIA; GOMES, DANIEL ; SILVA, ANTONIO T. e ; MUNIZ, RAFAEL O.R. ; MARTINS, MARCELO. Sensitivity assessment of fuel performance codes for loca accident scenario. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 22-27, 2017, Belo Horizonte, MG. Proceedings... Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear, 2017.

    Abstract: FRAPCON 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.

  • IPEN-DOC 24013

    GOMES, DANIEL S. ; SILVA, ANTONIO T. ; ABE, ALFREDO Y. ; MUNIZ, RAFAEL O.R. ; GIOVEDI, CLAUDIA. Simulation of accident-tolerant U3Si2 fuel using FRAPCON code. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 22-27, 2017, Belo Horizonte, MG. Proceedings... Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear, 2017.

    Abstract: The 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.

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A elaboração do projeto do RI do IPEN foi iniciado em novembro de 2013, colocado em operação interna em julho de 2014 e disponibilizado na Internet em junho de 2015. Utiliza o software livre Dspace, desenvolvido pelo Massachusetts Institute of Technology (MIT). Para descrição dos metadados adota o padrão Dublin Core. É compatível com o Protocolo de Arquivos Abertos (OAI) permitindo interoperabilidade com repositórios de âmbito nacional e internacional.

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