JOSE ANTONIO BATISTA DE SOUZA

JOSE ANTONIO BATISTA DE SOUZA

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Manufacturing high-uranium-loaded dispersion fuel plates in Brazil
2024 - DURAZZO, MICHELANGELO; SOUZA, JOSE A.B.; CARVALHO, ELITA F.U. de; RESTIVO, THOMAZ A.G.; GENEZINI, FREDERICO A.; LEAL NETO, RICARDO M.
The Nuclear and Energy Research Institute (IPEN-CNEN/SP) has developed and made available for routine production the technology for manufacturing dispersion-type fuel elements for research reactors. However, the fuel produced is limited to a uranium loading of 2.3 gU/cm3 (U3O8) or 3.0 gU/cm3 (U3Si2). To reduce Brazil’s dependence on foreign sources of Mo-99, the Brazilian government plans to construct a new research reactor, the 30 MW open pool Brazilian Multipurpose Reactor (RMB), which will mainly produce domestic Mo-99. Low-enriched uranium fuel will be used in the RMB, and increasing uranium loading will be important to increase the reactor core’s reactivity and fuel life. Uranium loadings of 3.2 gU/cm3 for the U3O8-Al and 4.8 gU/cm3 for the U3Si2-Al are considered the technological limit and have been well demonstrated worldwide. This work aimed to study the manufacturing process of these two highly uranium-loaded dispersion fuels and redefine current procedures. Additionally, UMo-Al dispersion fuel has been extensively studied globally and is likely to be the next commercially available technology. This new fuel utilizes a dispersion of UMo alloy with 7–10 wt% Mo, resulting in a uranium loading between 6 and 8 gU/cm3. We also studied this fuel type for potential use in the RMB research reactor. This work outlines the primary procedures for manufacturing these three types of fuels and the necessary adjustments to IPEN-CNEN/SP current technology. The manufacturing process proved to be well adapted to these new fuels, requiring only minor modifications. A comparison was made of the microstructures of fuel plate meat using three types of uranium compounds. The microstructures of U3Si2-Al and U10Mo-Al dispersions were found to be adequate, while that of U3O8-Al meat deviated significantly from the concept of an ideal dispersion.
Nickel electrodeposition in LEU metal foil annular targets to produce Mo-99
2022 - IANELLI, RICARDO F.; SALIBA-SILVA, ADONIS M.; TAKARA, ERIKI M.; GARCIA NETO, JOSE; SOUZA, JOSE A.B.; CARVALHO, ELITA F.U. de; DURAZZO, MICHELANGELO
The most used production route of Mo-99 is through the fission of U-235 in irradiation targets that are irradiated in research reactors. The annular target is a promisor design since it can incorporate high U-235 quantities, thus increasing the production yield of Mo-99. This target type uses a foil of uranium metal enveloped by a thin nickel foil that acts as a diffusion barrier. The process of uranium enveloping with nickel foil is today done manually. This operation risks the nickel foil breaking up during target assembling. In the present work, we studied the nickel electrodeposition over uranium metal foil surfaces to replace nickel foils. A pre-forming procedure of the uranium metal foil by calendering was developed to facilitate the assembling operation. The electrodeposition was done over the uranium foil pre-conformed in a tubular shape. An automated apparatus for electrodeposition of nickel in uranium tubular-shaped foil was developed. The results showed that the high nickel adherence to uranium metal depends on the proper activation of the uranium surface. Among the activation processes studied, the mechanical activation showed good adhesion of the nickel layer, with a loss of only 0.16% of uranium mass. Homogeneous and regular 12 μm thickness electrodeposited layers over the uranium metal were obtained. This work showed that the process could be used in continuous production technology, such as the production of irradiation targets.
Distribuição de vazão entre os canais de resfriamento do elemento combustível do IEA-R1
2022 - TORRES, WALMIR M.; UMBEHAUN, PEDRO E.; ANDRADE, DELVONEI A.; SOUZA, JOSE A.B.
Um elemento de combustível “dummy” instrumentado (DMPV-01) com as mesmas características geométricas de um elemento de combustível MTR foi projetado e construído para experimentos de medição de distribuição de vazão no núcleo do reator IEA-R1. Esse elemento instrumentado também foi usado para medir a distribuição de vazão entre os canais retangulares formados pelas placas do elemento combustível. Duas sondas com tomadas de pressão foram construídas e montadas dentro dos canais de escoamento para medir a queda de pressão, enquanto a velocidade de escoamento foi calculada usando uma equação de queda de pressão para canais fechados. Este trabalho apresenta o procedimento experimental e os resultados da medição da distribuição de vazão entre os canais de escoamento. Os resultados mostram que a vazão nos canais periféricos é de 10% a 15% menor que a vazão média. É importante conhecer a vazão nos canais periféricos devido a incertezas nos valores da vazão no canal aberto formado entre dois elementos combustíveis adjacentes. Essas vazões são responsáveis pelo resfriamento de placas externas do elemento combustível.
Manufacturing LEU-foil annular target in Brazil
2022 - DURAZZO, MICHELANGELO; SOUZA, JOSE A.B.; IANELLI, RICARDO F.; TAKARA, ERIKI M.; GARCIA NETO, JOSE S.; SALIBA-SILVA, ADONIS M.; CARVALHO, ELITA F.U. de
Molybdenum-99 is the most important isotope because its daughter isotope, technetium-99m, has been the most used medical radioisotope. The primary method used to produce Mo-99 derives from the fission of U-235 incorporated in so-called irradiation targets. Two routes are being developed to make Mo-99 by fissioning with low enriched uranium (LEU) fuel. The first adopts UAlx-Al dispersion plate targets. The second uses uranium metal foil annular targets. The significant advantage of uranium foil targets over UAlx-Al dispersion targets is the high density of uranium metal. This work presents the experience obtained in the development of the uranium metal annular target manufacturing steps. An innovative method to improve the procedure for assembling the uranium foil on the tubular target was presented. The experience attained will help the future production of Mo-99 in Brazil through the target irradiation in the Brazilian Multipurpose Reactor (RMB).
Kinects and factors on chemical dissolution of aluminum alloy AA6061 in NaOH alkaline media
2020 - TAKARA, E.M.; SOUZA, J.B. de; CARVALHO, E.F.U.; SILVA, A.S.
Nuclear Medicine is the Field of science that uses radioactive materials in order to diagnose and treat human body deceases. One of the most used radioisotopes for images diagnose purpose is the metastable technetium-99 (99mTc) because of its low decay half life (6 hours) and energy emission of 140keV that ensures low exposition time with the capacity of generating high quality images. The 99mTc is generated by the molibdenum-99(99Mo) radioactive decay during about 66 hours. The 99Mo is fabricated via nuclear fission of low encriched uranium (LEU) through plate irradiation targets (UAlx). The irradiation target cladding is made of Aluminum alloy AA6061 and its substrate is composed by 235U powder scattered in an AA1050 matrix. In general, studies are made targeting the prevention of corrosion mechanisms but the chemical dissolution in alkaline media, under hot cells, are one of the steps required for the post-processing methods of irradiation targets The time spent after irradiation is an important factor because the half life radioactive decay of the produced radioisotopes is relative short, then the procedures of dissolution, extraction, purify and distribution must be optimized in order to increase efficiency. This work presents a study of the factors impact involved on the chemical dissolution of the cladding aluminum alloys (temperature, NaOH solution concentration and dissolution time) as well as the kinects of the process associating it with the formation and destruction of oxides using electrochemical impedance spectroscopy (EIS) and scanning electron microscopy (SEM). It was found that the involved parameters contribute individually more effective and that there is no relevant association between the factors. Solution temperature showed to be the most influent factor following by exposition time. It was presented a equivalent circuit model which demonstrates the reaction kinects and the growing of passive layers that slow down the process before it turns up into a soluble phase.
Effects of Picture Frame Technique (PFT) on the corrosion behavior of 6061 aluminum alloy
2020 - MILAGRE, MARIANA X.; DONATUS, UYIME; MOGILI, NAGA V.; MACHADO, CARULINE S.C.; ARAUJO, JOAO V.S.; KLUMPP, RAFAEL E.; FERNANDES, STELA M.C.; SOUZA, JOSE A.B. de; COSTA, ISOLDA
The 6061 Al–Mg–Si alloy is used in nuclear fuel plates of nuclear research reactors which are fed with fuel in plate shapes. The production of these plates is based on the picture frame technique (PFT). The picture frame technique (PFT) is a manufacturing process for the fabrication of nuclear fuel plates where the nuclear fuel is encapsulated by Al alloy plates and thermomechanically processed to generate a set with reduced thickness. The effects of PFT on the corrosion resistance of the 6061 aluminum alloy were evaluated in this study by immersion and electrochemical tests in 0.005 mol L−1 NaCl solution. The results showed that the PFT fabrication process increases the corrosion resistance of the 6061 alloy in relation to the conventional 6061-T6, due phase dissolution and lower content of β’’ phase. Also, corrosion propagation gradually changes, with an increasing number of processing steps, from intergranular to intragranular corrosion attack.
Coatings for safe long term wet storage of spent Al-clad research reactor fuels
2015 - RAMANATHAN, L.V.; FERNANDES, S.M.C.; CORREA, O.V.; SOUZA, J.A. de; ANTUNES, R.A.; OLIVEIRA, M.C.L. de
Pitting corrosion of the aluminium cladding of spent research reactor (RR) fuels in wet storage has been observed and the use of conversion coatings to protect the cladding was proposed. A coating prepared by conventional chemical processing as opposed to electrochemical processing is the only option due to constraints related to the shape of the fuel and its high radioactivity. Hence, hydrotalcite (HTC) and boehmite were considered. This paper presents: (a) preparation of hydrotalcite (HTC) coatings from different baths followed by post-coating treatments; (b) corrosion behavior of HTC coated AA 6061 alloy; (c) results of field studies in which uncoated and HTC coated AA 6061 alloy coupons and plates, the latter assembled as a dummy fuel element, were exposed to the IEA-R1 reactor spent fuel basin for extended periods. The laboratory and field tests revealed marked improvements in the corrosion resistance of HTC coated specimens, coupons and plates. The mechanism of corrosion protection is presented.
Procedures for manufacturing an instrumented nuclear fuel element
2019 - DURAZZO, M.; UMBEHAUN, P.E.; TORRES, W.M.; SOUZA, J.A.B.; SILVA, D.G.; ANDRADE, D.A.
The IEA-R1 is an open pool research reactor that operated for many years at 2 MW. The reactor uses plate type fuel elements which are formed by assembling eighteen parallel fuel plates. During the years of reactor operation at 2 MW, thermohydraulic safety margins with respect to design limits were always very high. However, more intense oxidation on some external fuel plates was observed when the reactor power was increased to 5 MW. At this new power level, the safety margins are significantly reduced due to the increase of the heat flux on the plates. In order to measure, experimentally, the fuel plate temperature under operation, an instrumented fuel element was constructed to obtain temperature experimental data at various positions of one or more fuel plates in the fuel element. The manufacturing method is characterized by keeping the original fuel element design specifications. Type K stainless sheathed thermocouples are mounted into supports pads in unrestricted positions. During the fuel element assembling, the supports pads with the thermocouples are mechanically fixed by interference between two adjacent fuel plates. The thermocouple wires are directed through the space existing at the bottom of the mounting slot where the fuel plate is fixed to the side plates. The number of thermocouples installed is not restricted and depends only on adaptations that can be made on the mounting slots of the standard fuel element side plates. This work describes the manufacturing procedures for assembling such an instrumented fuel element.
Effect of processing parameters on hydrotalcite (HTC) coating microstructure and the corrosion behavior of HTC coated AA 6061 alloy
2018 - RAMANATHAN, L.V.; FERNANDES, S.C.; CORREA, O.V.; SOUZA, J.A.B.; ANTUNES, R.A.; LIMA, N.B.
Pitting corrosion of the aluminium cladding of spent nuclear fuels stored in light water pools has been observed. To prevent this, coating of the Al cladding with hydrotalcite (HTC) was proposed. This paper presents the effect of various processing parameters on HTC microstructure and the corrosion behavior of HTC coated AA 6061 specimens. The HTC coating from the high temperature nitrate bath was homogeneous, thicker and consisted of well-defined intersecting platelets than that formed from the room temperature carbonate bath. Electrochemical polarization measurements as well as long term exposure to aggressive aqueous media of HTC coated AA 6061 specimens revealed that specimens coated with HTC from the nitrate bath and further treated in a cerium salt solution were the most resistant to corrosion. The mechanism by which the HTC coating and cerium protect the Al alloy is discussed.
Thermal hydraulic analysis improvement for the IEA-R1 research reactor and fuel assembly design modification
2018 - UMBEHAUN, PEDRO E.; TORRES, WALMIR M.; SOUZA, JOSE A.B.; YAMAGUCHI, MITSUO; SILVA, ANTONIO T. e; MESQUITA, ROBERTO N. de; SCURO, NIKOLAS L.; ANDRADE, DELVONEI A. de
This paper presents the sequence of activities to improve the thermal hydraulic analysis of the IEA-R1 research reactor to operate in safe conditions after power upgrade from 2 to 5 MW and core size reduction from 30 to 24 fuel assemblies. A realistic analysis needs the knowledge of the actual operation conditions (heat flow, flow rates) beyond the geometric data and the uncertainties associated with manufacturing and measures. A dummy fuel assembly was designed and constructed to measure the actual flow rate through the core fuel assemblies and its pressure drop. First results showed that the flow distribution over the core is nearly uniform. Nevertheless, the values are below than the calculated ones and the core bypass flow rate is greater than those estimated previously. Based on this, several activities were performed to identify and reduce the bypass flow, such as reduction of the flow rate through the sample irradiators, closing some unnecessary secondary holes on the matrix plate, improvement in the primary flow rate system and better fit of the core components on the matrix plate. A sub-aquatic visual system was used as an important tool to detect some bypass flow path. After these modifications, the fuel assemblies flow rate increased about 13%. Additional tests using the dummy fuel assembly were carried out to measure the internal flow distribution among the rectangular channels. The results showed that the flow rate through the outer channels is 10% - 15% lower than the internal ones. The flow rate in the channel formed between two adjacent fuel assemblies is an estimated parameter and it is difficult to measure because this is an open channel. A new thermal hydraulic analysis of the outermost plates of the fuel assemblies takes into account all this information. Then, a fuel design modification was proposed with the reduction of 50% in the uranium quantity in the outermost fuel plates. In order to avoid the oxidation of the outermost plates by high temperature, low flow rate, a reduction of 50% in the uranium density in the same ones was shown to be adequate to solve the problem.