Eventos - Artigos: Recent submissions

  • IPEN-DOC 26783

    SANTOS, ADIMIR dos ; YAMAGUCHI, MITSUO ; FANARO, LEDA C.C.B. ; SANTOS, DIOGO F. dos ; SOUZA, GREGORIO S. de ; JUNQUEIRA, FERNANDO de C. ; SILVA, GRACIETE S. de A. e ; BELCHIOR JUNIOR, ANTONIO ; PRADO, ADELK de C. ; JOAO, THIAGO G.; ROSSI, PEDRO C.R.. New plate-type core of the IPEN/MB-01 research reactor facility for validation of RMB project. In: INTERNATIONAL CONFERENCE ON RESEARCH REACTORS: ADDRESSING CHALLENGES AND OPPORTUNITIES TO ENSURE EFFECTIVENESS AND SUSTAINABILITY, November 25-29, 2019, Buenos Aires, Argentina. Proceedings... Vienna, Austria: International Atomic Energy Agency, 2020.

    Abstract: The IPEN/MB-01 research reactor had its first criticality in November 1988 and, ever since, has been of major importance in Brazilian reactor physics researches, achieving international level for experiments comparison and validation (benchmarks). In this facility it is possible to build many different core configurations (i.e., rectangular, square and cylindrical), once versatility and flexibility were both taken into account on its initial project. The core is a fissile material assembly, inserted in a water tank, where the chain reaction is self-maintained and controlled at low power levels, so that, in normal operation, the feedback effects of temperatures are negligible. The core is intended for neutrons simulation of light water moderated reactors allowing the experimental verification of the calculation methods, reactor cell and mesh structures, control rods effectiveness, isothermal reactivity coefficients and core dynamics due to reactivity insertions. The first standard IPEN/MB-01 core had UO2 rod-type fuel, 4.3 % enriched in U-235 and using B4C and Ag-In-Cd rods for safety and control of the reactor. The facility is located at IPEN/CNEN-SP (Nuclear and Energy Research Institute), in Sao Paulo - Brazil. Within the scope of the new research reactor project, the Brazilian Multipurpose Reactor (RMB), it was designed a new critical configuration for the IPEN/MB-01. After thirty years of work, the rod-type fuels were replaced by plate-type fuels, in order to validate the RMB calculation methodologies, as well as the nuclear data libraries used. The RMB is an open pool-type reactor with maximum power of 30 MW, being the core a 5x5 configuration, consisting of 23 fuel elements, made of U3Si2-Al, having a medium density of 3.7 gU/cm3 and 19.75% enriched in U-235, and two positions available in the core for materials irradiation devices. The production of radioisotopes, silicon doping, neutron activation analysis, nuclear fuels and structural materials testing and the development of scientific and technological research using neutron beams are the main targets of the RMB enterprise. The new IPEN/MB-01 core has a 4×5 configuration, having 19 fuel elements, consisting of U3Si2-Al, 2.8 gU/cm³ and 19.75% enriched in U-235, plus one aluminum block. The IPEN/MB-01 new plate-type fuel assembly uses Cadmium wires as burnable poison, as the one used in RMB core for controlling the core power density and excess of reactivity during its operation. The core is also reflected by 4 boxes of heavy water (D2O), inserted in a moderator tank of light water. The maximum nominal power is 100 W and, for a safe operation, the critical assembly has both safety and auxiliaries’ systems. This paper presents a description of the new core and the principal neutronic parameters. The new core of the IPEN/MB-01 will be certainly a world class benchmark core for the core physics calculation of research reactors.

    Palavras-Chave: burnable poisons; control rod worths; cylindrical configuration; fuel elements; heavy water; neutron activation analysis; neutron beams; reactor cores; rmb reactor; uranium 235

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  • IPEN-DOC 26782

    FUNGARO, D.A. ; ROVANI, S. . Extração de sílica a partir das cinzas de resíduo da cana-de-açúcar em diferentes condições de extração alcalina. In: BALDOVI, ALDREW A. (Ed.); CONCEIÇÃO, ANA C.S. (Ed.); ANTENOR, ANANDA de O.G. (Ed.); CHYOSHI, BRUNA (Ed.); COSTA, DANILO O. da (Ed.); ANDRADE, HEDLLA M. (Ed.); FARIA, JULIA K. (Ed.); KOHATSU, MARCIO Y. (Ed.); MENDES, MARIANA E. (Ed.); COELHO, LUCIA, H.G. (Coord.); TAMBOSI, LEANDRO R. (Coord.) SIMPÓSIO DE CIÊNCIA E TECNOLOGIA AMBIENTAL, 1st, 2-3 de outubro, 2019, Santo André, SP. Anais... Santo André, SP: Universidade Federal do ABC, 2019. p. 239-244.

    Abstract: Uma amostra de cinzas geradas durante a queima de resíduos de cana-de-açúcar foi triturada com NaOH sólido variando-se a relação cinzas: NaOH (m/m). As misturas foram submetidas ao processo de fusão a 450 oC por 1 h. Após o resfriamento, a sílica da massa fundida foi lixiviada para a fase líquida na forma de silicato de sódio. Seguindo a separação do sólido e do líquido, partículas de sílica foram precipitadas pela adição de ácido sulfúrico até pH 7 à solução de silicato de sódio. O rendimento das amostras mostrou-se constante e uma pureza de 89% foi alcançada. As amostras de cinzas e de sílica produzidas foram caracterizadas por difração de raios-X, fluorescência de raios-X e análise elementar.

    Palavras-Chave: sugar cane; biomass; ashes; silica

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  • IPEN-DOC 26781

    URBANI, G.L.; FRANCO, M.K.K.D. ; YOKAICHIYA, F. ; VICENTE, R. . Aplicação da química de radiação à questões tecnológicas do cimento relacionadas ao desenvolvimento de repositórios de rejeitos radioativos do modelo borehole. In: BALDOVI, ALDREW A. (Ed.); CONCEIÇÃO, ANA C.S. (Ed.); ANTENOR, ANANDA de O.G. (Ed.); CHYOSHI, BRUNA (Ed.); COSTA, DANILO O. da (Ed.); ANDRADE, HEDLLA M. (Ed.); FARIA, JULIA K. (Ed.); KOHATSU, MARCIO Y. (Ed.); MENDES, MARIANA E. (Ed.); COELHO, LUCIA, H.G. (Coord.); TAMBOSI, LEANDRO R. (Coord.) SIMPÓSIO DE CIÊNCIA E TECNOLOGIA AMBIENTAL, 1st, 2-3 de outubro, 2019, Santo André, SP. Anais... Santo André, SP: Universidade Federal do ABC, 2019. p. 226-231.

    Abstract: Resíduos radioativos são usualmente descartados em repositórios do tipo borehole ou de superfície. Por questões de segurança, devem ser depositados em tambores e cobertos por cimento na sua destinação final. A radiação gama proveniente dos resíduos radioativos interage com a água livre da pasta de cimento e causa o fenômeno da radiólise. Essa interação que decompõe a água da pasta de cimento é estudada pela academia científica e ainda não é um consenso se a mesma afeta a resistência do concreto ou não. Para um melhor entendimento dessa questão, nesse estudo um modelo teórico simples é sugerido para quantificar a porcentagem da água que sofre radiólise e se essa perda é suficiente para afetar a resistência da barreira de cimento. O resultado indica que a quantidade de água perdida neste processo não é suficiente para diminuir a resistência do concreto. O modelo proposto foi aplicado para verificar os resultados experimentais, utilizando as condições iniciais expostas na literatura existente, com o propósito comparação e discussão sobre o fenômeno da radiólise.

    Palavras-Chave: radioactive waste storage; storage facilities; portland cement; portland cement; boreholes; radiolysis

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  • IPEN-DOC 26770

    GOMES, DANIEL de S. . Study of thoria-urania fuel during accidents. In: INTERNATIONAL CONGRESS OF MECHANICAL ENGINEERING, 25th, October 20-25, 2019, Uberlândia, MG. Proceedings... Rio de Janeiro, RJ: Associação Brasileira de Engenharia e Ciências Mecânicas, 2019.

    Abstract: In this investigation, nuclear fuel based on mixed ceramic oxides, using (Th–U)O2 as nuclear fuel and zirconium-based alloy as cladding, was simulated. This strategic configuration can achieve improved safety margins because of an enhanced set of thermal and mechanical properties. The Experimental Breeder Reactor built in the 1950s in Idaho introduced the concept that a reactor can generate more fissile material than it consumes. The thorium fuels have a lower cost and should decrease weapon-grade plutonium compared with conventional fuel, UO2. The nuclear characteristic of thorium-232 or U-238 can make a converter into U-233 or Pu-239. However, using thoria fuels can avoid weapon proliferation by reducing plutonium, and it also should reduce radionuclides such as (Np, Am, Cm). This study uses an optimized composition of Th-75% wt and U-25% wt with an enrichment of 19.5%. We studied the behavior using the fuel licensing codes FRAPCON and FRAPTRAN, including many adaptations for the mixed composition choice. The results prove that thoria–urania fuel has a higher performance than pure uranium dioxide fuel during accidents.

    Palavras-Chave: thorium; uranium; loss of coolant; accidents; nuclear fuels

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  • IPEN-DOC 26769

    COSTADELLE, EWERTON L. ; FIGUEIREDO, NEY G.F. ; BARRETO, ROGERIO L. ; ALMEIDA, GISELE F.C. ; COUTO, ANTONIO A. . Creep-testing machine retrofit for Ti-6Al-4V alloy study. In: INTERNATIONAL CONGRESS OF MECHANICAL ENGINEERING, 25th, October 20-25, 2019, Uberlândia, MG. Proceedings... Rio de Janeiro, RJ: Associação Brasileira de Engenharia e Ciências Mecânicas, 2019.

    Abstract: The Nuclear and Energy Research Institute received two Creep-Testing Machine frame donated from Technological Institute of Aeronautics in nonfunctional mode. These apparatuses has arrived with structure, lever arm and tube furnace. However, they came with no strain gage and no temperature controller. In order to back the machinery to active mode, this work developed a Creep-Testing Machine modernization suite. In fact, this study developed a new PID controller hardware, based in an Arduino open platform. In addition, it used a LVDT developed in Brazil, by Technological Research Institute, to capturing the specimen strain. After the modernization suite implementation, it evaluated the creep strain-rate of the Ti6Al4V alloy at 873 K in 319 MPa. Moreover, compared the results of the same lot material tested in another two ones apparatuses. This open technology was able to maintain the specimen temperature in the set point, getting and saving the test results in a text file and it got results close to the most modern equipment.

    Palavras-Chave: retrofitting; titanium; creep

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  • IPEN-DOC 26768

    CASTRO, PEDRO A.A. de ; ZEZELL, DENISE M. . Infrared Spectroscopy evaluation of burn wound healing: semi-quantitative study. In: QUINCY BROWN, J. (Ed.); VAN LEEUWEN, TON G. (Ed.) EUROPEAN CONFERENCES ON BIOMEDICAL OPTICS, June 23-27, 2019, Munich, Germany. Proceedings... Bellingham, WA, USA: SPIE, 2019. p. 1107304-1 - 1107304-4. (Proceedings of SPIE-OSA Vol. 11073, Clinical and Preclinical Optical Diagnostics II). DOI: 10.1117/12.2527051

    Abstract: Wound healing is a biological response in order to recover the tissue stability after injury. The impaired healing by thirddegree, when the damage achieves the major part of dermis, is defined in four sequential and overlapping phases: Inflammation, transition, proliferative and maturative1. The role of biochemical cascade associated in each phase are still not fully understood, thus systematic evaluations tests are crucial. In fact, the gold standard to interrogate the molecular signature of wound healing is concern on immunohistochemical analysis. This approach tends to be laborious, timeconsuming and require multiple assays2. Since Fourier transform infrared spectroscopy (FTIR) has been demonstrated in other studies to provide molecular change report upon biological samples, the present study aims to estimate the feasibility of FTIR to discriminate healthy and burned skin throughout wound stages.

    Palavras-Chave: burns; wounds; infrared spectra

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  • IPEN-DOC 26767

    LOPES, MONICA S.; MOTA, CLAUDIA C.B.O.; PEREIRA, DAISA L. ; AMARAL, MARCELLO M.; ZEZELL, DENISE M. ; GOMES, ANDERSON S.L.. Effect of Nd:YAG laser and aluminum oxide sandblasting preconditioning on lingual enamel: brackets shear bond strength and morphological characterization. In: WOJTKOWSKI, MACIEJ (Ed.); BOPPART, STEPHEN A. (Ed.); OH, WANG-YUHL (Ed.) EUROPEAN CONFERENCES ON BIOMEDICAL OPTICS, June 23-27, 2019, Munich, Germany. Proceedings... Bellingham, WA, USA: SPIE, 2019. p. 1107822-1 - 1107822-3. (Proceedings of SPIE-OSA Vol. 11078, Optical Coherence Imaging Techniques and Imaging in Scattering Media III). DOI: 10.1117/12.2527030

    Abstract: It is known that Nd:YAG laser radiation on dental structure can increase enamel resistance to demineralization; however, few studies report its impact in orthodontics. This study aimed to verify the interaction of Nd:YAG laser and aluminum oxide sandblasting (Al2O3) as preconditioning treatment of lingual brackets bonding to quantify the shear bond strength (SBS) and to characterize the enamel after different surface preconditioning techniques. Thirty-five bovines’ incisors were divided in 5 groups (n=7), according to the preconditioning. All groups had the lingual brackets bonded with Transbond XT adhesive according to manufacturer’ protocol, and shear bonded after 72h. Samples were analyzed by Optical Coherence Tomography (OCT) and Scanning Eletronic Microscope (SEM) to verify enamel alterations. Optical attenuation coefficient (α) was measured before any preconditioning (T0) and after preconditioning surface (T1) for each group. Statistics analysis ANOVA-test was used, followed by Post Hoc Tukey for shear bond strength data, and Kruskal Wallis and post hoc Dunn test for α data. SEM and OCT qualitative analysis showed the melting effect with laser irradiation and enamel crystal surface disorganization with sandblasting in T1 and after shear bond. All groups were adequate for SBS and the statistical differences (p=0.0034) among groups showed better results for groups with techniques association (laser and Al2O3 used in this or in reverse order) and the highest SBS when laser was used after. The Al2O3 removes part of melting effect. The α had statistical difference (p= 0.0124) among groups, increasing with laser and Al2O3 isolated and decrease with techniques association.

    Palavras-Chave: tomography; coherent radiation; aluminium oxides; oral cavity; dentistry

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  • IPEN-DOC 26766

    LIMA, CASSIO A. ; CORREA, LUCIANA; BYRNE, HUGH J.; ZEZELL, DENISE M. . Assessing the spectrochemical signatures of skin components using FTIR microspectroscopy. In: QUINCY BROWN, J. (Ed.); VAN LEEUWEN, TON G. (Ed.) EUROPEAN CONFERENCES ON BIOMEDICAL OPTICS, June 23-27, 2019, Munich, Germany. Proceedings... Bellingham, WA, USA: SPIE, 2019. p. 110730S-1 - 110730S-3. (Proceedings of SPIE-OSA Vol. 11073, Clinical and Preclinical Optical Diagnostics II). DOI: 10.1117/12.2527137

    Abstract: Fourier Transform Infrared (FTIR) spectroscopy is a label-free analytical technique used to evaluate the chemical profile of a sample based on its molecular vibrations. The potential dermatological applications of FTIR spectroscopy has been well demonstrated over the past decades through many proof-of-concept studies evaluating cancerous and non-cancerous cutaneous diseases. Considering that the correctly identification of skin components plays an important role in the study of cutaneous diseases, the present study aims to evaluate the spectrochemical signatures of dermis and epidermis based on the pixels of a FTIR hyperspectral image collected from healthy skin.

    Palavras-Chave: infrared spectrometers; skin; epithelium; cluster analysis

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  • IPEN-DOC 26762

    SIERRA, JULIAN H.; CARVALHO, DANIEL O.; SAMAD, RICARDO E. ; RANGEL, RICARDO C.; ALAYO, MARCO I.. Analysis and measurement of the non-linear refractive index of SiOxNy using pedestal waveguides. In: SYMPOSIUM ON MICROELECTRONICS TECHNOLOGY AND DEVICES, 34th, August 26-30, 2019, São Paulo, SP. Proceedings... Piscataway, NJ, USA: IEEE, 2019. DOI: 10.1109/SBMicro.2019.8919392

    Abstract: In this work, the non-linear refractive index (n2) of silicon oxynitride (SiOxNy) is determined, obtaining a value for this material of n2 = 2.11×10-19 m2/W. The results demonstrate that this material has interesting properties for the development of non-linear optical devices. The paper presents in detail the waveguide fabrication process using the pedestal technique, which allows using different materials since it does not require etching to define the sidewalls of the waveguides. We show the results of the measurement of the n2 employing the non-linear optical phenomena of Self-Phase Modulation (SPM).

    Palavras-Chave: optical equipment; photons; silicon nitrides; silicon oxides; microelectronics

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  • IPEN-DOC 26761

    OLIVEIRA, L.S. ; CORREA, O.V. ; BENTO, R.T. ; PILLIS, M.F. . Síntese de alumina anódica nanoporosa em liga de Alumínio AA 1050. In: CONGRESSO BRASILEIRO DE QUÍMICA, 59., 5-8 de novembro, 2019, João Pessoa, PB. Anais... Rio de Janeiro, RJ: Associação Brasileira de Química, 2019.

    Abstract: Nos últimos anos, alumina anódica porosa tem atraído interesse devido ao arranjo regular dos nanoporos, à possibilidade de controle de seu diâmetro, e grande área de superfície específica. Neste trabalho foram obtidas estruturas de alumina anódica porosa por meio da anodização em duas etapas de uma liga de alumínio AA 1050 em eletrólito de H2SO4. As amostras foram anodizadas durante 4 e 8 h. A técnica de microscopia eletrônica de varredura com canhão de emissão de elétrons foi utilizada para a avaliação morfológica da superfície. Os nanoporos formados após 4h de anodização apresentaram distribuição homogênea na superfície da liga e diâmetro médio de 25 nm, enquanto que após 8h de anodização os nanoporos apresentaram-se com formato irregular e distribuição não homogênea.

    Palavras-Chave: anodization; aluminium alloys; porous materials

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  • IPEN-DOC 26760

    BORAZANIAN, T.C.F. ; SZURKALO, M. ; CORREA, O.V. ; BENTO, R.T. ; SANTOS, T.F. ; COTINHO, S.P. ; PILLIS, M.F. . Revestimentos de TiO2 para preservação de superfícies cimentícias. In: CONGRESSO BRASILEIRO DE QUÍMICA, 59., 5-8 de novembro, 2019, João Pessoa, PB. Anais... Rio de Janeiro, RJ: Associação Brasileira de Química, 2019.

    Abstract: Este trabalho tem como objetivo avaliar o desempenho autolimpante de filmes fotocatalíticos de TiO2 aplicados sobre argamassa cimentícia usada como revestimento externo de vedações verticais das edificações. Os filmes foram sintetizados pelo método sol-gel e depositados por spray coating, à temperatura ambiente, sobre substratos de argamassa cimentícia. Foram testadas amostras com uma e duas camadas de deposição de filme e foram realizados testes de autolimpeza. As análises apontam que as amostras com aplicação de duas camadas de filme de TiO2 exibem degradação mais eficiente do corante azul de metileno, após 3 ciclos de 48 horas sob radiação UV, o que sugere sua aplicação promissora para preservação e manutenção de superfícies de revestimentos externos utilizados na construção civil.

    Palavras-Chave: titanium oxides; construction; cleaning; films

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  • IPEN-DOC 26759

    OLIVEIRA, E.C. de ; CORREA, O.V. ; BENTO, R.T. ; COTINHO, S.P. ; SANTOS, T.F. dos ; PILLIS, M.F. . Caracterização morfológica de filmes de TiO2 dopados com nitrogênio crescidos por MOCVD. In: CONGRESSO BRASILEIRO DE QUÍMICA, 59., 5-8 de novembro, 2019, João Pessoa, PB. Anais... Rio de Janeiro, RJ: Associação Brasileira de Química, 2019.

    Abstract: O método de deposição química de organometálicos em fase vapor (MOCVD) foi utilizado para o crescimento de filmes de dióxido de titânio (TiO2) e TiO2 dopado com nitrogênio. Os filmes foram crescidos a 400 °C sobre substratos de vidro borossilicato. Isopropóxido de titânio IV foi utilizado como precursor de titânio e de oxigênio, e amônia (NH3) como fonte de nitrogênio. Análises por microscopia de força atômica (AFM) mostraram que ambos os filmes apresentaram grãos bem definidos e arredondados. Todos os filmes são formados apenas pela fase cristalina anatase. Os resultados mostraram que a dopagem com nitrogênio resultou em uma diminuição no tamanho médio de grão e na rugosidade superficial.

    Palavras-Chave: titanium oxides; organometallic compounds; chemical vapor deposition; doped materials; nitrogen

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  • IPEN-DOC 26758

    ROVANI, S. ; CARVALHO, F. ; SANTOS, J.; RAMOS, N.; MORANDI, M.; SALDANHA, M.; FUNGARO, D. . Caracterização físico-química das propriedades de cinzas de cana-de-açúcar de diferentes usinas brasileiras. In: CONGRESSO BRASILEIRO DE QUÍMICA, 59., 5-8 de novembro, 2019, João Pessoa, PB. Anais... Rio de Janeiro, RJ: Associação Brasileira de Química, 2019.

    Abstract: O interesse em transformar biomassa em novas fontes de energia e novos materiais vem fomentando estudos em todo o mundo. As cinzas resultantes da queima de palha e bagaço de cana, por exemplo, é um material rico em sílica. No presente estudo, amostras de cinzas de resíduos de cana-de-açúcar foram coletadas nas usinas de Cerradinho (Chapadão do Céu, GO), Iracema (Iracemápolis, SP) e Guaíra (Guaíra, SP), localizadas em regiões onde os solos são classificados como Latossolos. As cinzas foram caracterizadas pelas técnicas de XRD, EDX, MEV, espectroscopia no IV e DTG. A maioria das amostras de cinzas apresentou sílica como principal constituinte (42-69%) e a sílica na forma cristalina em todas. Diferenças significativas nas concentrações de outros elementos foram observadas.

    Palavras-Chave: sugar cane; ashes; silica; biomass

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  • IPEN-DOC 26757

    BENTO, R.T. ; CORREA, O.V. ; COTINHO, S.P. ; SANTOS, T.F. dos ; PILLIS, M.F. . Avaliação do efeito da morfologia e da espessura de filmes de TiO2 na degradação do corante alaranjado de metila. In: CONGRESSO BRASILEIRO DE QUÍMICA, 59., 5-8 de novembro, 2019, João Pessoa, PB. Anais... Rio de Janeiro, RJ: Associação Brasileira de Química, 2019.

    Abstract: Filmes de TiO2 com diferentes espessuras foram crescidos por MOCVD sobre vidro borossilicato a 400°C. Os efeitos das características morfológicas e da espessura dos filmes sobre sua atividade fotocatalítica foram avaliados a partir da degradação do corante alaranjado de metila sob luz UVA. Os resultados apontaram a existência de um valor ideal de espessura, no qual o catalisador exibe o melhor desempenho fotocatalítico. O filme de TiO2 com espessura de 470 nm exibiu o melhor comportamento, com uma eficiência de 65,3 % em 5 horas de uso. O filme apresentou uma elevada estabilidade fotocatalítica, após diversos ciclos de utilização, o que permite a sua aplicação prática no tratamento de água a partir de um método verde, e com alta durabilidade e eficiência

    Palavras-Chave: titanium oxides; organometallic compounds; chemical vapor deposition; photocatalysis

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  • IPEN-DOC 26756

    PIERETTI, E.; BORAZANIAN, T. ; CORREA, O. ; PILLIS, M. ; ANTUNES, R.. Análise eletroquímica de um biomaterial revestido por TiO2. In: CONGRESSO BRASILEIRO DE QUÍMICA, 59., 5-8 de novembro, 2019, João Pessoa, PB. Anais... Rio de Janeiro, RJ: Associação Brasileira de Química, 2019.

    Abstract: O presente trabalho avaliou, por técnicas eletroquímicas, superfícies do aço inoxidável austenítico ISO 5832-1 recobertas com filmes de TiO2, sintetizados pelo método sol-gel. As técnicas eletroquímicas utilizadas foram: monitoramento de potencial de corrosão em circuito aberto em função do tempo de imersão em solução de Ringer, que simula os fluidos corpóreos; espectroscopia de impedância eletroquímica e cálculo de densidades de dopantes no filme passivo por meio da abordagem de Mott-Schottky. Amostras deste mesmo biomaterial sem recobrimento foram analisadas para fins de comparação. Os resultados indicaram diminuição da susceptibilidade à corrosão localizada nas amostras recobertas por TiO2, devido ao caráter aderente e protetor dos filmes depositados nas superfícies deste biomaterial.

    Palavras-Chave: stainless steels; films; titanium oxides; surgical materials; biological recovery

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  • IPEN-DOC 26755

    PEREIRA, DAISA L. ; DEL VALLE, MATHEUS; GOMES, GABRIELA V. ; ZEZELL, DENISE M. ; ANA, PATRICIA A.. Optical properties of bovine dentin when irradiated by Nd:YAG and a black dentifrice aimed at treating dentin erosion. In: COSTA-FELIX, RODRIGO (Ed.); MACHADO, JOÃO C. (Ed.); ALVARENGA, ANDRÉ V. (Ed.) BRAZILIAN CONGRESS ON BIOMEDICAL ENGINEERING, 26th, October 21-25, 2018, Armação de Buzios, RJ. Proceedings... Singapore: Springer Nature Singapore, 2019. p. 847-850. (IFMBE Proceedings 70/2). DOI: 10.1007/978-981-13-2517-5_131

    Abstract: Dental erosion has been extensively studied as a risk factor for tooth loss or injure, and the early diagnosis of lesions is essential for avoiding greater damages. Optical Coherence Tomography (OCT) is a potential tool for early diagnosis of demineralization. In this study, this technique was used to analyze the optical changes of dentin samples irradiated with Nd:YAG laser using a black dentifrice as photoabsorber, then submitted to an erosive cycling. 75 slabs of bovine root dentin were randomized into 5 groups: G1—untreated; G2—treated with acidulated phosphate fluoride gel (APF-gel, [F-] = 1.23%, pH = 3.3–3.9); G3—irradiated with Nd: YAG laser (100 μs, 1064 nm, 0.6 W, 10 Hz) without photoabsorber; G4—irradiated with Nd:YAG laser using a coal paste as photoabsorber; G5—irradiated with Nd: YAG laser using a black dentifrice as photoabsorber. All samples were submitted to a 3-day erosive demineralization (Citric acid 1%, pH = 3.6, 5 min, 2 /day) under agitation, and remineralization (artificial saliva, pH = 7, 120 min) cycling. The samples were evaluated by OCT before treatments (baseline), after treatments and after erosive cycling. Optical attenuation coefficient (μ) was calculated using a Matlab routine, and the statistical analysis was performed (a = 0.05). It was observed a significant decrease on μ values after all treatments. Also, the μ values decreased after erosive cycling, except for the groups G3 and G5. It was concluded that OCT technique is capable to distinguish among sound, treated and demineralized dentin. As well, the black paste was efficient to act as a photoabsorber, helping the Nd:YAG laser to decrease dentin erosion.

    Palavras-Chave: dentistry; dentin; caries; teeth; cattle; lasers; optical properties; infrared spectra

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  • IPEN-DOC 26754

    FERREIRA, ELIZABETE dos S.; PRATES, ILKA T.K.; SANTOS JUNIOR, SERGIO L.M. dos; DEL VALLE, MATHEUS; ZEZELL, DENISE M. ; ANA, PATRICIA A.. In vitro study of Er,Cr:YSGG laser effects when used for the prevention of dentin demineralization. In: COSTA-FELIX, RODRIGO (Ed.); MACHADO, JOÃO C. (Ed.); ALVARENGA, ANDRÉ V. (Ed.) BRAZILIAN CONGRESS ON BIOMEDICAL ENGINEERING, 26th, October 21-25, 2018, Armação de Buzios, RJ. Proceedings... Singapore: Springer Nature Singapore, 2019. p. 825-829. (IFMBE Proceedings 70/2). DOI: 10.1007/978-981-13-2517-5_127

    Abstract: Erbium lasers can be used to prevent dental caries, which has a high prevalence in the worldwide population. However, effective irradiation parameters for root dentin have not yet been determined using the Er,Cr:YSGG laser. Objective: this study evaluated the chemical, morphological and optical effects of Er,Cr:YSGG laser on root dentin when aimed at preventing root caries. Methodology: 75 bovine root dentin slabs were randomly distributed in 5 groups: G1-untreated; G2-treated with acidulated phosphate fluoride gel (APF-gel, [F−]= 1.23%); G3-Er,Cr: YSGG laser irradiation (2.78 μm, 60 μs, 6 J/cm2, 8,67 mJ/pulse, 0.25 W); G4-Laser irradiation + APF-gel application; G5-APF-gel application + Laser irradiation. The chemical and morphological evaluations were performed using Fourier transformed infrared spectroscopy and scanning electron microscopy, respectively. Afterwards, the samples were submitted to an 8-day pH-cycling model and the optical attenuation coefficient was evaluated by optical coherence tomography. The statistical analysis was performed considering the level of significance of 5%. Laser irradiation alone does not alter the dentin composition, but the previous application of APF-gel followed by laser irradiation significantly decreased the content of m3m4 carbonate of dentin. This treatment also promoted greater morphological alterations, such as ablation of the surface, when compared to the treatments alone. After demineralization, this treatment also presented the highest optical attenuation coefficient value when compared to the other treatments, indicating less demineralization of the samples. Conclusion: Er,Cr: YSGG laser presents potential for use in prevention of root dentin demineralization, and is more efficient when preceded by the application of APF.

    Palavras-Chave: dentistry; dentin; caries; erbium; roots; teeth; lasers; biological radiation effects; demineralization

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  • IPEN-DOC 26753

    PANESI, RICARDO ; BERUSKI, OTAVIO ; KORKISCHKO, IVAN ; OLIVEIRA NETO, ALMIR ; SANTIAGO, ELISABETE . Modeling and parametric analysis of PEM fuel cells using computational fluid dynamics. In: INTERNATIONAL CONGRESS OF MECHANICAL ENGINEERING, 25th, October 20-25, 2019, Uberlândia, MG. Proceedings... Rio de Janeiro, RJ: Associação Brasileira de Engenharia e Ciências Mecânicas, 2019.

    Abstract: This paper presents a parametric investigation of PEMFC electrochemical models employing computational fluid dynamics (CFD) technique and aims to determine the relative importance of each parameter on the modeling results. A compatible and systematic mathematical model is developed in order to study the effect of these parameters. The model is applied to an isothermal, steady state an single phase to observe the main results by a polarization curve. The results compare well with the experimental polarization data obtained at 80 ºC for ohmic and activation regions. The best match with the experimental data is obtained when the specific active surface area of the catalyst layer is 700 cm2/mg and electrolyte conductivity of 8 S/m.

    Palavras-Chave: computerized simulation; fluid mechanics; proton exchange membrane fuel cells; fuel cells; numerical analysis; simulation

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  • IPEN-DOC 26557

    GOMES, JOAO P.C. ; SOUZA, KATIA R. de; PEDRIALI-MORAES, CARLA A.; SEO, EMILIA S.M. . Sabonete com base auto-emulsionante polimérica A/O com potencial repelente. In: CONGRESSO LATINO-AMERICANO E IBÉRICO DE QUÍMICOS COSMÉTICOS, 24., 21-23 de maio, 2019, São Paulo, SP. Anais... 2019. p. 1-11.

    Abstract: O uso de repelentes é uma proteção individual, a qual é de extrema importância para evitar os surtos promovidos pela febre amarela, dengue, febre Chikungunya e o vírus Zika. Os repelentes mais eficazes são os tópicos convencionais com ativos sintéticos e naturais. Como alternativa mais efetiva existem os repelentes tópicos convencionais, que podem ser sintéticos ou naturais. Para minimizar os riscos de esquecimento ou do uso incorreto dos repelentes tópicos convencionais (como em forma de creme, gel ou líquido), este trabalho teve a proposta alternativa do desenvolvimento do sabonete com poliamida 3, sendo uma base sabonete com potencial de repelência. Sua vantagem, além da praticidade, é assegurar a proteção do usuário desde o banho formando um filme resistente à água e assim, evitar a não proteção em razão do esquecimento do uso dos cremes ou sprays. O filme formado sobre a pele é oriundo da junção da Polyamide 3 Resin com os emolientes: Propylene Glycol Dicaprylate/ Dicaprate, Isopropyl Myristate, Isostearyl Alcohol e Isostearyl Isostearate. Os ativos Cymbopogon nardus e Ethyl butylacetylaminopropionate promovem a repelência. A base dos sabonetes moldado e gel cremoso foi mais estável quando a poliamida é submetida a alta temperatura para a incorporação dos emolientes a sua estrutura. Os ativos estudados tiveram melhor eficácia quando o sabonete esteve com o pH na faixa 5,5 à 6,3. As formulações base e com o ativo óleo essencial de citronela apresentaram mínima alteração no pH.

    Palavras-Chave: other organic compounds; soaps; emulsification; organic polymers

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  • IPEN-DOC 26549

    BAPTISTA, A.; NUNEZ, S.C.; MARTIN, A.A.; RIBEIRO, M.S. . Targets of photodyamic inactivation in fungal cells. In: HASAN, TAYYABA (Ed.) INTERNATIONAL PHOTODYNAMIC ASSOCIATION WORLD CONGRESS, 17th, June 28 - July 4, 2019, Cambridge, Massachusetts, USA. Proceedings... Bellingham, WA, USA: SPIE, 2019. p. 11070BY-1 - 11070BY-6. (Proceedings SPIE 11070). DOI: 10.1117/12.2537128

    Abstract: Photodynamic inactivation (PDI) has been reported to be effective to eradicate a wide variety of pathogens, including antimicrobial-resistant microorganisms. However, there are conflicting reports in the literature about the effect of growth phase on the susceptibility to PDI. The aim of this study was to identify the potential molecular targets of PDI on Candida albicans in exponential growth phase after PDI mediated by methylene blue (50μM) and exposure to a 660nm-LED (P=360mW). For this task, scanning electron microscopy (SEM) and Fourier transform infrared spectroscopy (FT-IR) techniques were employed. Pre-irradiation time was set at 10min and exposure time was 15 min delivering a radiant exposure of 162 J/cm2 on a 24-well plate of about 2 cm2. Morphological analysis revealed cell damage after PDI. FT-IR predominantly showed degradation of functional groups related to C-O of deoxyribose; C-C of DNA; C-O stretching vibration of C-OH group of ribose-RNA; P-O stretching modes from the phosphodiester groups of nucleic acids; C=C, C=N, C=O, N=H proteins and amides. Previous studies from our group had demonstrated different targets on the same cells but in stationary growth phase. Therefore, we can conclude that PDI promoted damage to intracellular structures in fungal cells at exponential-phase growth and information on the susceptibility of different growth phases to PDI can be of great importance for the development of treatment strategies that would lead to inactivation of fungal cells in all possible phases of growth in a way that would turn the clinical PDI treatment effective and predictable.

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  • IPEN-DOC 26536

    RIBEIRO, MARTHA S. ; SABINO, CAETANO P.; NUNEZ, SILVIA C.. Antimicrobial photodynamic therapy: from basis to clinical applications. In: HASAN, TAYYABA (Ed.) INTERNATIONAL PHOTODYNAMIC ASSOCIATION WORLD CONGRESS, 17th, June 28 - July 4, 2019, Cambridge, Massachusetts, USA. Proceedings... Bellingham, WA, USA: SPIE, 2019. p. 1107048-1 - 1107048-9. (Proceedings SPIE 11070). DOI: 10.1117/12.2527918

    Abstract: Antimicrobial photodynamic therapy (APDT) combines the use of light with a photosensitizer (PS) and oxygen to kill microbial cells. Even though this technique was first reported in the beginning of the 20th century, APDT never took off as antimicrobial chemotherapy did. However, microbial resistance to chemotherapy is currently expanding in faster rates than drug discovery. Therefore, introduction of therapeutic alternatives that bypass mechanisms of drug resistance now presents an urgent status. Fortunately, the scientific and technological development related to APDT made it far more feasible for mainstream clinical applications. Our research group has been working on mechanisms and applications of APDT for almost 20 years. We have already reported that successful APDT results depend on a number of factors, such as PS and light parameters, cell type, and oxygen abundance, among others. We have also demonstrated that APDT is an effective adjuvant in endodontics and periodontics and can be a non-invasive treatment for caries, candidiasis and cutaneous leishmaniasis. In Veterinary Medicine, we have reported effective treatment for penguin pododermatitis, snake stomatitis and dog otitis. This presentation will give an integrated perspective from the basic APDT mechanisms, preclinical and clinical trials to protocol optimization and future perspectives.

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  • IPEN-DOC 26752

    ARQUINTO, JULIANA ; SILVA, LEONARDO G. de A. e ; ESPER, FABIO J.; ZACHARIAS, JANICE M.; FILHO, MARCOS M.O.. Produção de bioplastico utilizando amido da semente de jaca (Artocarpus heterophyllus Lam) / Bioplastic production using jackfruit seed starch (Artocarpus heterophyllus Lam). In: CONGRESSO BRASILEIRO DE POLÍMEROS, 15., 27-31 de outubro, 2019, Bento Gonçalves-RS. Anais... São Carlos, SP: Associação Brasileira de Polímeros, 2019. p. 1627-1632.

    Abstract: A intensão deste projeto é apresentar um estudo sobre a influência da palha de milho na propriedade de resistência a tração até ruptura de formulações de bioplástico tendo como base o amido extraído da semente da jaca juntamente com a utilização da glicerina e da trietaloamina como agentes plastificantes. Este bioplástico é um material que em relação ao plástico oriundo dos derivados de petróleo, denota um menor grau de impacto ao meio ambiente após seu descarte. Para isto, o trabalho consistiu inicialmente da extração do amido da semente, com posterior caracterização por FTIR e análise quantitativa do teor de amido no produto extraído. A etapa posterior consistiu da produção das amostras de bioplástico utilizando os agentes plastificantes em uma concentração fixa e variando-se a quantidade de palha de milho na formulação. Por fim, as amostras de bioplástico obtidas foram submetidas ao teste de tração até ruptura para análise de resistência do filme.

    Palavras-Chave: plastics; fruits; composite materials; seeds; starch; maize

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  • IPEN-DOC 26749

    VILLANI, D. ; RODRIGUES JUNIOR, O. ; CAMPOS, L.L. . Study on electronic equilibrium of 137Cs gamma radiation for 3D printed phantoms using OSL dosimetry. In: CONGRESSO BRASILEIRO DE METROLOGIA, 10.; CONGRESSO INTERNACIONAL DE METROLOGIA ELÉTRICA, 13.; CONGRESSO INTERNACIONAL DE METROLOGIA MECÂNICA, 5.; CONGRESSO BRASILEIRO DE METROLOGIA DAS RADIAÇÕES IONIZANTES, 6.; WORKSHOP DA REDE DE METROLOGIA QUÍMICA DO INMETRO, 4.; CONGRESSO BRASILEIRO DE METROLOGIA ÓPTICA, 3., 24-27 de novembro, 2019, Florianópolis, SC. Anais... Rio de Janeiro, RJ: Sociedade Brasileira de Metrologia, 2019.

    Abstract: With the popularization of 3D printing technologies, it is now possible to develop patient specific simulators and various other accessories using this technology in medical physics and dosimetry. This work aims to evaluate the electronic equilibrium of 3D printed phantoms using PLA and ABS filaments compared to PMMA for 137Cs gamma rays using OSL dosimetry. A Landauer microStar ii commercial OSL system were commissioned and it was used nanoDot dosimeters. Phantom plates with 2.5, 3.0 and 5.0 mm thickness were used to obtain electronic equilibrium for 137Cs gamma rays. Measurements were compared with PMMA measurements at standard conditions. Results show that measurements with ABS and PLA thicknesses of 2.5 and 3.0 mm presents dosimetry results within irradiation uncertainty. More accuracy is obtained using 3.0 mm for both PLA and ABS phantoms, with differences in less than 0.5%. It can be concluded that PLA and ABS 3D phantom plates has similar properties of PMMA for 137Cs gamma rays dosimetry and can be used for developing dosimetry accessories for this energy photon beam.

    Palavras-Chave: phantoms; 3d printing; gamma radiation; photoluminescence; dosimetry

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  • IPEN-DOC 26748

    ASSEMANY, L.P.F. ; RODRIGUES JUNIOR, O. ; POTIENS, M.P.A. . Reuse of 3D printed materials for dosimetry purposes. In: CONGRESSO BRASILEIRO DE METROLOGIA, 10.; CONGRESSO INTERNACIONAL DE METROLOGIA ELÉTRICA, 13.; CONGRESSO INTERNACIONAL DE METROLOGIA MECÂNICA, 5.; CONGRESSO BRASILEIRO DE METROLOGIA DAS RADIAÇÕES IONIZANTES, 6.; WORKSHOP DA REDE DE METROLOGIA QUÍMICA DO INMETRO, 4.; CONGRESSO BRASILEIRO DE METROLOGIA ÓPTICA, 3., 24-27 de novembro, 2019, Florianópolis, SC. Anais... Rio de Janeiro, RJ: Sociedade Brasileira de Metrologia, 2019.

    Abstract: 3D printing technology has been a great ally of the medical industry due to it allows the obtaining of anatomical structures such as custom prostheses, implants and surgery planning simulators for the most several applications. There are in the market, several types of filaments used for 3D printing, being the most used thermoplastics Acrylonitrile Butadiene Styrene (ABS) and Polylactic Acid (PLA). Within the modeling and printing process, tests are made with different printing parameters and often part of the test material is discarded. The objective of this work was to study a methodology for recycling discarded materials printed in 3D printer for use in characterization studies for dosimetry purposes.

    Palavras-Chave: anatomy; dosimetry; medicine; polymers; recycling; screen printing; technology utilization

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  • IPEN-DOC 26747

    SILVA, E. ; SANTOS, L.R. ; ASSEMANY, L.P.F. ; POTIENS, M.P.A. . Evaluation of the behavior of a 180cc ionization chamber under different environmental conditions. In: CONGRESSO BRASILEIRO DE METROLOGIA, 10.; CONGRESSO INTERNACIONAL DE METROLOGIA ELÉTRICA, 13.; CONGRESSO INTERNACIONAL DE METROLOGIA MECÂNICA, 5.; CONGRESSO BRASILEIRO DE METROLOGIA DAS RADIAÇÕES IONIZANTES, 6.; WORKSHOP DA REDE DE METROLOGIA QUÍMICA DO INMETRO, 4.; CONGRESSO BRASILEIRO DE METROLOGIA ÓPTICA, 3., 24-27 de novembro, 2019, Florianópolis, SC. Anais... Rio de Janeiro, RJ: Sociedade Brasileira de Metrologia, 2019.

    Abstract: The use of ionizing radiation for medical purposes has been a major advance for society due to the numerous possibilities of use for the treatment and diagnosis of diseases. On the other hand, knowledge about the damage caused by the biological effects of ionizing radiation requires continuous improvement of diagnostic radiology quality control. Radiation detector equipment is used to measure radiation levels emitted from natural or artificial sources. For convenience and accuracy, among the most widely used detectors are ionization chambers. Especially outdoors, weather factors can affect the behavior of these detectors at the time of measurement, but Brazilian law recommends only calibrating these measuring instruments in a traceable laboratory every two years to ensure their reliability. The objective of this work was to evaluate the performance of an ionization chamber used in radioprotection measurements in diagnostic radiology equipment, considering climatic variations in different regions of Brazil. For this, a system was developed to simulate the environmental conditions found for the temperature and humidity parameters at the moment of the clinical measurements, allowing to estimate the influence of these factors on the obtained values.

    Palavras-Chave: ionization chambers; nuclear medicine; performance testing; quality control; radiation metrology; radiation protection; radiology

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  • IPEN-DOC 26746

    ALMEIDA, J.S. ; VILLANI, D. ; POTIENS, M.P.A. ; WILLEGAIGON, J.. Dosimetric characterization of 3D printed for 137Cs gamma rays. In: CONGRESSO BRASILEIRO DE METROLOGIA, 10.; CONGRESSO INTERNACIONAL DE METROLOGIA ELÉTRICA, 13.; CONGRESSO INTERNACIONAL DE METROLOGIA MECÂNICA, 5.; CONGRESSO BRASILEIRO DE METROLOGIA DAS RADIAÇÕES IONIZANTES, 6.; WORKSHOP DA REDE DE METROLOGIA QUÍMICA DO INMETRO, 4.; CONGRESSO BRASILEIRO DE METROLOGIA ÓPTICA, 3., 24-27 de novembro, 2019, Florianópolis, SC. Anais... Rio de Janeiro, RJ: Sociedade Brasileira de Metrologia, 2019.

    Abstract: The aim this paper is characterize materials for 3D printed with different infill percentages for 137Cs gamma rays. The RAISE 3D PRO2 printer was used to print PLA and ABS plates. Using a 137Cs source, the attenuation coefficient was obtained by the transmission method and results compared with PMMA. The readings were performed by a Radcal ionization chamber, model 10X6-6. The results of attenuation coefficients show that the PLA filament demonstrated a equivalent behavior to PMMA. The PLA plates exhibits an increase in radiation transmission when reduces the infilling, and ABS printing achieved same results for all infills.

    Palavras-Chave: 3d printing; computer-aided fabrication; gamma radiation; cesium 137; phantoms

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  • IPEN-DOC 26745

    MARTINS, E.W. ; KUAHARA, L.T. ; POTIENS, M.P.A. . Development of an "in situ" calibration methodology to activity meters. In: CONGRESSO BRASILEIRO DE METROLOGIA, 10.; CONGRESSO INTERNACIONAL DE METROLOGIA ELÉTRICA, 13.; CONGRESSO INTERNACIONAL DE METROLOGIA MECÂNICA, 5.; CONGRESSO BRASILEIRO DE METROLOGIA DAS RADIAÇÕES IONIZANTES, 6.; WORKSHOP DA REDE DE METROLOGIA QUÍMICA DO INMETRO, 4.; CONGRESSO BRASILEIRO DE METROLOGIA ÓPTICA, 3., 24-27 de novembro, 2019, Florianópolis, SC. Anais... Rio de Janeiro, RJ: Sociedade Brasileira de Metrologia, 2019.

    Abstract: The performance of a safety and efficient practice of a nuclear medicine service depends, among other factors, on a complete quality control program, especially in the case of the radionuclide activity measuring instrument, the activimeter. Several factors may influence the accuracy of the measurements performed with an activimeter, and the largest sources of errors are related to the types of containers that contain radiopharmaceuticals (eg, thickness, size and volume). A complete quality control program should include the calibration of all measurement instruments used in the procedure. However, in Brazil, the actual standard that establishes the requirements of radiological protection for nuclear medicine services (NMS), does not include the calibration of the activimeter. Considering that these instruments, for various reasons, are difficult to remove for sending to a Calibration Service, the purpose of this work is to develop a methodology for activimeter calibration that can be applied "in situ" to the most used radiopharmaceutical, 99mTc.

    Palavras-Chave: activity meters; calibration; legislation; quality control; radiation protection; technetium 99

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  • IPEN-DOC 26744

    VILLANI, D. ; SAVI, M. ; ANDRADE, M.A.B.; CAMPOS, L.L. ; POTIENS, M.P.A. . Characterization of ABS + W and ABS + Bi 3D printing filaments attenuation for different photon beams. In: CONGRESSO BRASILEIRO DE METROLOGIA, 10.; CONGRESSO INTERNACIONAL DE METROLOGIA ELÉTRICA, 13.; CONGRESSO INTERNACIONAL DE METROLOGIA MECÂNICA, 5.; CONGRESSO BRASILEIRO DE METROLOGIA DAS RADIAÇÕES IONIZANTES, 6.; WORKSHOP DA REDE DE METROLOGIA QUÍMICA DO INMETRO, 4.; CONGRESSO BRASILEIRO DE METROLOGIA ÓPTICA, 3., 24-27 de novembro, 2019, Florianópolis, SC. Anais... Rio de Janeiro, RJ: Sociedade Brasileira de Metrologia, 2019.

    Abstract: 3D printing techniques and materials have become widely available in the last couple of decades and remains a hot topic of study as new materials can lead to new applications. This study aims to evaluate the attenuation behaviour of GMASS over photon beams ranging from 29.7 up to 661.7keV, comparing with pure ABS and using theoretical data of pure lead as reference. It was used the transmission method to obtain experimental attenuation coefficients to all materials and theoretical data. HVL and TVL calculations were also performed. Results show that ABS+W has higher attenuation than ABS+Bi and pure ABS. Using the lead theoretical reference data it can be concluded that although ABS+Bi and ABS+W attenuates less than pure lead, the 3D printing filaments can be used to create shielding tolls depending on radiation energy and application.

    Palavras-Chave: attenuation; cesium 137; dosimetry; energy dependence; filaments; photon beams; polymers; radiations; shielding

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  • IPEN-DOC 26743

    BUENO, LETICIA K. ; RODRIGUES JUNIOR, ORLANDO ; POTIENS, MARIA da P.A. . Avaliação da atenuação de invólucros produzidos em impressora 3D para medidas com calibrador de dose. In: CONGRESSO BRASILEIRO DE METROLOGIA, 10.; CONGRESSO INTERNACIONAL DE METROLOGIA ELÉTRICA, 13.; CONGRESSO INTERNACIONAL DE METROLOGIA MECÂNICA, 5.; CONGRESSO BRASILEIRO DE METROLOGIA DAS RADIAÇÕES IONIZANTES, 6.; WORKSHOP DA REDE DE METROLOGIA QUÍMICA DO INMETRO, 4.; CONGRESSO BRASILEIRO DE METROLOGIA ÓPTICA, 3., 24-27 de novembro, 2019, Florianópolis, SC. Anais... Rio de Janeiro, RJ: Sociedade Brasileira de Metrologia, 2019.

    Abstract: Devido ao aumento de procedimentos realizados nos Serviços de Medicina Nuclear (SMN) tornou-se cada vez mais inevitável a preocupação com o perfeito funcionamento dos calibradores de dose utilizados diariamente. Ainda assim, a prática segura, eficiente e eficaz do uso do equipamento envolve a integração de vários processos. O objetivo deste trabalho é o projeto e desenvolvimento de novos invólucros por meio da prototipação utilizando uma impressora 3D. Os materiais escolhidos foram o PLA e o ABS. Foram realizados testes de precisão e exatidão variando os parâmetros de impressão e as dimensões do invólucro. Os resultados mostraram que é possível customizar os invólucros melhorando os resultados e reduzindo as incertezas no controle de qualidade desses equipamentos.

    Palavras-Chave: calibration; computer-aided design; dosimetry; nuclear medicine; packaging; radiation doses; radioisotopes

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  • IPEN-DOC 26725

    MARTOS, LUIS G.C.; BRASCHI, GIOVANI F.; CARNEIRO, MARCELO B.; MACHADO, IZABEL F.; BARBOSA, PATRICIA A.; ROSSI, WAGNER de . Avaliação de elemento cerâmico em gradação funcional / Evaluation of ceramic element functionally graded. In: CONGRESSO BRASILEIRO DE ENGENHARIA DE FABRICAÇÃO, 10., 5-7 de agosto, 2019, São Carlos, SP. Anais... Rio de Janeiro, RJ: Associação Brasileira de Engenharia e Ciências Mecânicas, 2019.

    Palavras-Chave: sintering; microhardness; particle size; ceramics; density

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  • IPEN-DOC 26603

    TODO, ALBERTO S. ; CARDOSO, JOAQUIM C.S. ; RODRIGUES JUNIOR, ORLANDO . Analysis of the in vivo monitoring program at IPEN in the last 14 years. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 93-98.

    Abstract: This paper presents the results of the in vivo monitoring analysis for Occupationally Exposed Individuals that handle unsealed sources at Nuclear and Energy Research Institute (IPEN). The facilities are: the Radiopharmacy Center, the Cyclotron, the IEA-R1 Reactor and Research Laboratories. The in vivo monitoring program is carried out in a whole-body counter for radionuclides emitting gamma rays with energy above 100 keV. This system is equipped with 8x4 and 3x3 inch NaI(Tl) detector, for whole body and thyroid measurements, respectively. The objective of this work is to analyze the results of the internal monitoring program according to the dose received by the Occupationally Exposed Individuals from 2005 to 2018. During this period about 6,000 measurements were accomplished. The radionuclides that presented measured values above the detection limit of the system were: 131I, 99Mo, 99mTc, 153Sm, 177Lu, 111In, 192Ir, 125I, 123I, 181Hf, 203Hg, 67Ga, 18F, 51Cr, 201Tl. These measurements have amounted less than 6.9% of the total whole-body monitoring’s performed in this period. Among these radionuclides, 131I, 99mTc, 125I and 18F have contributed with 69% of all measurements above the limit of detection, but most dose results were below the recording level under installations normal operating conditions. Regarding to the radionuclides that have presented doses above the recording level we can mention the 131I, 67Ga, 111In that occurred in small unexpected situations. The results shown by this analysis give a good support to the internal individual monitoring program implemented by the radioprotection service in these facilities.

    Palavras-Chave: effective radiation doses; gamma radiation; in vivo; internal irradiation; nai detectors; occupational exposure; radiation monitoring; radiation protection; radioisotopes; thyroid; unsealed sources; whole-body counters; brazilian cnen

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  • IPEN-DOC 26548

    BORAZANIAN, TATYANA C.F. ; SZURKALO, MARGARIDA ; CORREA, OLANDIR V. ; BENTO, RODRIGO T. ; PILLIS, MARINA F. . Revestimentos de TiO2 para a preservação de superfícies arquitetônicas. In: CONGRESSO BRASILEIRO DE CATALISE, 20., 1-5 setembro, 2019, São Paulo, SP. Anais... 2019.

    Abstract: A utilização do dióxido de titânio (TiO2) tem sido amplamente estudada para proteção de elementos arquitetônicos e de revestimentos externos utilizados na construção civil, a fim de preservar melhor seu aspecto visual e minimizar a necessidade constante de limpeza e manutenção decorridas da deposição de partículas de poluentes existentes na atmosfera. Este trabalho tem como objetivo avaliar o desempenho autolimpante de filmes fotocatalíticos de TiO2 aplicados em superfícies de materiais comumente empregados na Arquitetura.

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  • IPEN-DOC 26547

    BENTO, RODRIGO T. ; CORREA, OLANDIR V. ; PILLIS, MARINA F. . Reativação e reutilização de fotocatalisadores de TiO2 dopados com enxofre em baixa temperatura. In: CONGRESSO BRASILEIRO DE CATALISE, 20., 1-5 setembro, 2019, São Paulo, SP. Anais... 2019.

    Abstract: O presente estudo avaliou a possibilidade de reutilização fotocatalítica dos filmes de dióxido de titânio (TiO2) dopados com enxofre em baixa temperatura. Os filmes foram crescidos por deposição química de organometálicos em fase vapor (MOCVD) a 400°C. A dopagem com enxofre foi realizada a 50°C por um processo semelhante ao utilizado na dessulfuração do sulfeto de hidrogênio (H2S). O comportamento fotocatalítico e a durabilidade dos filmes foram medidos a partir da degradação do corante alaranjado de metila sob luz visível por vários ciclos. Os filmes são formados apenas pela fase cristalina anatase. Os resultados demonstraram que não houve modificações estruturais ou diferenças significativas na morfologia dos filmes após a sua utilização. Os filmes de TiO2 dopados com enxofre apresentaram uma excelente atividade fotocatalítica, com uma eficiência de 72,1% sob luz visível. Os experimentos de durabilidade sugerem que, mesmo com a impregnação de corante na superfície do catalisador, os filmes de TiO2 dopados apresentaram boa estabilidade fotocatalítica após diversas horas de uso, o que permite sua aplicação prática no tratamento e purificação da água sob luz solar com elevada eficiência.

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  • IPEN-DOC 26546

    ALENCAR, CATARINE S.L. ; PAIVA, ANA R.N. ; VAZ, JORGE M. ; SPINACE, ESTEVAM V. . Preparação de nanopartículas de cobre e ouro suportadas em TiO2 para uso como catalisador na oxidação preferencial de CO em misturas ricas em hidrogênio (CO-PROX). In: CONGRESSO BRASILEIRO DE CATALISE, 20., 1-5 setembro, 2019, São Paulo, SP. Anais... 2019.

    Abstract: Os catalisadores Au/TiO2 têm apresentado boa atividade e seletividade para a reação de oxidação preferencial de monóxido de carbono em misturas ricas em hidrogênio (CO-PROX). É proposto um catalisador contendo os metais Au e Cu (CuAu/TiO2) que será preparado por meio de redução química de forma simultânea de ambos os metais, utilizando borohidreto de sódio como agente redutor. Realizou-se também a síntese de catalisadores monometálicos de Cu/TiO2 e Au/TiO2 sob as mesmas condições e os resultados foram comparados. Os catalisadores foram caracterizados por Difração de Raios X (DRX), Energia Dispersiva de Raios X (EDX), Microscopia Eletrônica de Transmissão (MET) e Redução por temperatura programada (TPR). O catalisador CuAu/TiO2 apresentou melhor atividade catalítica para a reação CO-PROX se comparado aos seus respectivos catalisadores monometálicos.

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  • IPEN-DOC 26545

    QUEIROZ, CARLA M.S. ; MACHADO, ARTHUR P. ; PAIVA, ANA R.N. ; VAZ, JORGE M. ; SPINACE, ESTEVAM V. . Preparação de catalisadores Pt/CeO2 promovidos por Fe e Sn via método de redução por álcool para a oxidação preferencial de CO em misturas ricas em hidrogênio (PROX-CO). In: CONGRESSO BRASILEIRO DE CATALISE, 20., 1-5 setembro, 2019, São Paulo, SP. Anais... 2019.

    Abstract: A reação de oxidação preferencial do CO (PROX-CO) é considerada uma alternativa eficiente e econômica para remoção de CO presente em correntes de H2 que são empregadas para a produção de energia limpa e sustentável via tecnologia de células a combustível. Isto porque, a reação de PROX-CO catalisada é capaz de reduzir a concentração deste contaminante para níveis menores que 50 ppm, evitando assim o envenenamento e consequente desativação dos eletrodos da célula. Os catalisadores de Pt têm se mostrado bastante promissores para emprego na reação de PROX-CO, apresentado altas conversões e seletividades, numa ampla faixa de temperatura. Assim, este trabalho tem o objetivo de estudar o desempenho de catalisadores de Pt/CeO2 promovidos com óxidos de Fe ou Sn destinados à reação de PROX-CO. Os sólidos foram sintetizados pelo método de redução por álcool e caracterizados pelas técnicas de EDX, DRX e MET. Os desempenhos catalíticos foram conduzidos sob pressão atmosférica e em temperaturas variando entre 50 °C e 200 °C. Os resultados revelam que os catalisadores de Pt suportados em céria apresentaram máxima conversão de CO e seletividade em CO2 a 50 °C.

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  • IPEN-DOC 26541

    EMILIOZZI, CAROLINE Z.S.; MENEZES, MARIO O. de ; MOREIRA, EDSON G. . Investigation of the parameters affecting patients waiting time in the radiotherapy treatment by using algorithms to evaluate electronic health records. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 1659-1668.

    Abstract: Cancer is the second leading cause of death worldwide and Radiotherapy (RT) is an important modality in the treatment of these patients, which consists in deploying ionising radiation to destroy or damage cancer cells. With this growing global burden, demand for RT has been increasing continuously and supply-demand imbalances have become a major concern. The reason is that delays in radiotherapy can affect the outcome by permitting local proliferation of clonogenic cells and spread of the cancer beyond the treatment volume. Studies show a common cause of anxiety for radiotherapy patients is the fact that they do not know how long they will have to wait for treatment to start. In this study, we analyze the data of electronic health records to attempt to provide a better understanding of the problem and provide an initial estimate of radiotherapy patient’s waiting time. The data for this project comes from a subset of MOSAIQ, a relational database system developed by Elekta and used as an electronic health record system by the Radiation Oncology Department at the Hospital das Clínicas de São Paulo (HCFMUSP). The dataset consists of real historical data collected between January 2016 and December 2018. Visual Basic for application (VBA) and RSTUDIO Software were used to extract and analyze the data. Our work goal is to investigate a set of factors and verify their influence on patient waiting time. Factors as diagnosis, patient’s age, priority of the diagnosis, and the season in which treatment planning has initiated may reveal crucial information about overall efficiency and guide us to improve clinical procedures and practices.

    Palavras-Chave: algorithms; delayed radiation effects; neoplasms; patients; radiotherapy; records management; schedules; time delay

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  • IPEN-DOC 26539

    CASTRO, MAYSA C. de ; SILVA, NATALIA F. da ; CALDAS, LINDA V. E. . Dosimetric tests of an extrapolation chamber in standard computed tomography beams. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 116-120.

    Abstract: Computed tomography (CT) diagnostic exams are responsible for the highest dose values to the patients. Therefore, the radiation doses in this procedure must be accurate. For the dosimetry of CT beams, the radiation detector is usually a pencil-type ionization chamber. This type of dosimeter presents a uniform response to the incident radiation beam from all angles, which makes it suitable for such equipment since the X-ray tube executes a circular movement around the table during irradiation. However, there is no primary standard system for this kind of radiation beam yet. In order to search for a CT primary standard, an extrapolation chamber built at the Calibration Laboratory (LCI) of the Instituto de Pesquisas Energéticas e Nucleares (IPEN) was tested. An extrapolation chamber is a parallel-plate ionization chamber that allows the variation of its sensitive air volume. This chamber was used previously for low-energy radiation beams and showed results within the international recommended limits. The aim of this work is to perform some characterization tests (saturation curve, polarity effect, ion collection efficiency and linearity of response) considering the chamber depth of 1.25 mm in the radiation qualities for computed tomography beams at the LCI. The results showed to be within the international recommended limits.

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  • IPEN-DOC 26535

    SILVA, PAULO S.C. da ; CAMPOS, MARCIA P. de ; REIS, GUILHERME de L. . Radon concentrations on the nuclear and radioactive instalations of nuclear reactor center – CRPQ/IPEN. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 105-115.

    Abstract: Nuclear and radioactive workers are normally exposed to dose resulting from their day by day activities. Besides that, the ubiquitous radon distribution can also contribute for the exposure rates. The radionuclide 222Rn is a noble gas belonging to the uranium series and its indoor concentration in the air depend on exhalation from surrounding soil and on exhalation from building materials. Radon emanating from porous building materials may achieve large relevance in areas with high uranium concentrations and areas with limited ventilation. The objective of this study was to evaluate the 222Rn concentrations in the radiochemistry and radiometric laboratories and in the reactor nuclear building of the Nuclear Reactor Center (CERPq) located in the Nuclear and Energy Research Institute (IPEN). Measurements were done by using a Radon Gas Monitor, model RAD7, produced by Durridge Company equipped with a solid state alpha detector and a passive method, with SSNTDs placed within small diffusion chambers, as detectors square pieces (2.5 cm × 2.5 cm) of CR 39 foils were used. The CR 39 detectors were etched in KOH 30% solution at 80 °C for 5.5 h in a constant temperature bath. After etching, the detectors were washed, dried, and scanned under a Carl Zeiss microscope to obtain the track density measurements. The activity concentrations varied from 52 to 103 Bq m 3 for the measured areas in CERPq. These values are in accordance with what is stablished by the World Health Organization for safe environments of 100 Bq m 3.

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  • IPEN-DOC 26517

    CAPPUZZELLO, F.; AGODI, C.; ACOSTA, L.; AMADOR-VALENZUELA, P.; AUERBACH, N.; BAREA, J.; BELLONE, J.I.; BELMONT, D.; BIJKER, R.; BONANNO, D.; BORELLO-LEWIN, T.; BOZTOSUN, I.; BRANCHINA, V.; BRASOLIN, S.; BRISCHETTO, G.; BRUNASSO, O.; BURRELLO, S.; CALABRESE, S.; CALABRETTA, L.; CALVO, D.; CAPIROSSI, V.; CARBONE, D.; CAVALLARO, M.; CHEN, R.; CIRALDO, I.; CHAVEZ LOMELI, E.R.; COLONNA, M.; D'AGOSTINO, G.; DJAPO, H.; DE GERONIMO, G.; DELAUNAY, F.; DESHMUKH, N.; FARIA, P.N. de; ESPEJEL, R.; FERRARESI, C.; FERREIRA, J.L.; FERRETTI, J.; FINOCCHIARO, P.; FIRAT, S.; FISICHELLA, M.; FLORES, A.; FOTI, A.; GALLO, G.; GARCIA-TECOCOATZI, H.; GONGORA, B.; HACISALIHOGLU, A.; HAZAR, S.; HUERTA, A.; KOTILA, J.; KUCUK, Y.; IAZZI, F.; LANZALONE, G.; LA VIA, F.; LAY, J.A.; LENSKE, H.; LINARES, R.; LONGHITANO, F.; LO PRESTI, D.; LUBIAN, J.; MA, J.; MARIN-LAMBARRI, D.; MARTINEZ, S.; MAS, J.; MEDINA, N.H.; MENDES, D.R.; MEREU, P.; MORALLES, M. ; OLIVEIRA, J.R.B.; ORDONEZ, C.; PAKOU, A.; PANDOLA, L.; PETRASCU, H.; PIETRALLA, N.; PINNA, F.; REITO, S.; REZA, G.; RIES, P.; RIFUGGIATO, D.; RODRIGUES, M.R.D.; RUSSO, A.D.; RUSSO, G.; SANDOVAL, S.; SANTOPINTO, E.; SANTOS, R.B.B.; SGOUROS, O.; SILVEIRA, M.A.G. da; SOLAKCI, S.O.; SOULIOTIS, G.; SOUKERAS, V.; SPATAFORA, A.; TORRESI, D.; TUDISCO, S.; VSEVOLODOVNA, R.I.M.; VARGAS, H.; VEGA, G.; WANG, J.S.; WERNER, V.; YANG, Y.Y.; YILDIRIN, A.; ZAGATTO, V.A.B.. The NUMEN project @ LNS: status and perspectives. In: MARUYAMA, REINA (Ed.) SYMMETRIES AND ORDER: ALGEBRAIC METHODS IN MANY BODY SYSTEMS, October 5-6, 2018, Connecticut, USA. Proceedings... Melville, NY, USA: AIP Publishing, 2019. p. 030003-1 - 030003-6. (AIP Conference Proceedings, 2150). DOI: 10.1063/1.5124592

    Abstract: The NUMEN project aims at accessing experimentally driven information on Nuclear Matrix Elements (NME) involved in the half-life of the neutrinoless double beta decay (0υββ), by high-accuracy measurements of the cross sections of Heavy Ion (HI) induced Double Charge Exchange (DCE) reactions. Particular interest is given to the (18O,18Ne) and (20Ne,20O) reactions as tools for β+β+ and β-β- decays, respectively. First evidence about the possibility to get quantitative information about NME from experiments is found for both kind of reactions. In the experiments, performed at INFN - Laboratory Nazionali del Sud (LNS) in Catania, the beams are accelerated by the Superconducting Cyclotron (CS) and the reaction products are detected by the MAGNEX magnetic spectrometer. The measured cross sections are challengingly low, limiting the present exploration to few selected isotopes of interest in the context of typically low-yield experimental runs. A major upgrade of the LNS facility is foreseen in order to significantly increase the experimental yield, thus making feasible a systematic study of all the cases of interest. Frontiers technologies are going to be developed, to this purpose, for the accelerator and the detection systems. In parallel, advanced theoretical models are developed aiming at extracting the nuclear structure information from the measured cross sections.

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  • IPEN-DOC 26534

    SANTOS, GIVANILDO A. dos; GUIMARÃES, MARCOS de A.; AGUIAR, HERBERT C.G. de; NASCIMENTO, MAURICIO S.; SANTOS, VINICIUS T. dos; SILVA, MARCIO R. da; COUTO, ANTONIO A. ; BATALHA, GILMAR F.. Análise do acabamento superficial na usinagem de liga de bronze-alumínio-níquel sem utilização de fluido de corte / Analysis of surface finish in machining of bronze-aluminum-nickel alloy without using cutting fluid. In: CONGRESSO BRASILEIRO DE ENGENHARIA DE FABRICAÇÃO, 10., 5-7 de agosto, 2019, São Carlos, SP. Anais... Rio de Janeiro, RJ: Associação Brasileira de Engenharia e Ciências Mecânicas, 2019.

    Abstract: As ligas bronze-alumínio-níquel têm usinabilidade de 20 a 40% comparada com o latão de corte livre, de forma que os parâmetros de corte destas ligas podem diferir bastante dos utilizados para outras ligas de cobre e a ausência de contaminação por fluidos de corte possibilita um maior valor na venda dos cavacos, encorajando a utilização de usinagem sem refrigeração. Neste trabalho foi avaliado o comportamento quanto ao acabamento superficial na usinagem da liga bronze-alumínio-níquel CuAl10Ni5Fe5 com pastilha de metal duro em diferentes velocidades de corte, sem utilização de fluido de corte (refrigerante). Os resultados mostraram menor valor e maior estabilidade da rugosidade em condições de velocidades de corte mais altas.

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  • IPEN-DOC 26363

    ABE, ALFREDO Y. ; MELO, CAIO; GIOVEDI, CLAUDIA; SILVA, ANTONIO T. . Modification of TRANSURANUS fuel performance code in the ATF framework. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5036-5045.

    Abstract: 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.

    Palavras-Chave: accident-tolerant nuclear fuels; aluminium alloys; chromium alloys; cladding; comparative evaluations; fuel rods; iron alloys; loss of coolant; performance; t codes; zirconium alloys

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  • IPEN-DOC 26357

    LIMA, LEONARDO S.; MELO, CAIO; FARIA, DANILO P.; BERRETA, JOSE; ABATI, AMANDA ; GIOVEDI, CLAUDIA . Analysis of stresses acting on the internal and external surfaces of fuel rod of a pressurized water reactor using computational simulation. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4950-4961.

    Abstract: During operation of a Pressurized Water Reactor, the cladding of the fuel rod is subjected to various loads, such as: temperature, internal pressure, and external pressure which generate dimensional and geometric variations in the cladding tube. In the fuel rod, at the operating temperature, the internal pressure comes from the initial pre-pressurizing with Helium gas and the release of fission gases by the UO2 pellets during the irradiation. The external pressure is assumed to be the same as that of the coolant. In this paper, it was proposed the study of a mathematical model for computational simulation using the Finite Element Method to calculate and analyze the mechanical stresses acting on the internal and external surfaces of the fuel rod, adopting the normal operating condition, at 0 W of power. The boundary conditions, such as temperature and pressure profile, come from a modified version of a fuel performance code, considering as cladding material an iron-based alloy (austenitic stainless steel). The fuel rod was modeled and simulated using the Solidworks and ANSYS softwares, respectively. The values of the stresses acting on the cladding tube obtained by simulation were compared to the values obtained by analytical calculation. Then, it was checked the consistency of the adopted mathematical model, in order to ensure the reliability of the computational simulation as a tool to evaluate the stresses acting on the internal and external surfaces of the fuel rod under a PWR environment.

    Palavras-Chave: a codes; austenitic steels; boundary conditions; cladding; computerized simulation; finite element method; fuel rods; iron alloys; pwr type reactors; s codes; stresses

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  • IPEN-DOC 26356

    GIOVEDI, CLAUDIA; MELO, CAIO; ABE, ALFREDO Y. ; SILVA, ANTONIO T. ; MARTINS, MARCELO R.. Fuel performance of iron-based alloy cladding using modified TRANSURANUS code. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4943-4949.

    Abstract: 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.

    Palavras-Chave: cladding; comparative evaluations; computerized simulation; fuel rods; iron alloys; nuclear fuels; performance; pwr type reactors; stainless steel-348; steady-state conditions; t codes; zircaloy 4

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  • IPEN-DOC 26350

    OLIVA, AMAURY M. ; ALVES FILHO, HERMES; BARROS, RICARDO C.; CURBELO, JESUS P.. The spectral deterministic method applied to nêutron fixed-source discrete ordinates problems in X, Y-geometry for multigroup calculations. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4702-4716.

    Abstract: A new approach for the development of a coarse-mesh numerical spectral nodal method is presented in this paper. This method, referred to as the Spectral Deterministic Method { Constant Nodal (SDM{CN), is based on a spectral analysis of the multigroup X,Y-Geometry, linearly anisotropic scattering neutron transport equations in discrete ordinates ( SN)formulation for xed-source calculations in non-multiplying media. In this paper we present typical model problems to illustrate the accuracy and the e ciency for coarse-mesh energy multigroup SN calculations of the SDM-CN method. The numerical results obtained are compared with the traditional ne-mesh Diamond Di erence (DD) method and the results obtained by DOT{II and TWOTRAN codes. The numerical results are also compared with the spectral nodal method, spectral Green's function (SGF).

    Palavras-Chave: boltzmann equation; comparative evaluations; comparative evaluations; computerized simulation; d codes; discrete ordinate method; finite difference method; multigroup theory; neutron transport; nodal expansion method; scattering; t codes

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  • IPEN-DOC 26223

    DOURADO, NELSON X. ; OMI, NELSON M. ; SOMESSARI, SAMIR L. ; GENEZINI, FREDERICO A. ; FEHER, ANSELMO ; NAPOLITANO, CELIA M. ; AMBIEL, JOSE J. ; CALVO, WILSON A.P. . Preliminary studies on the development of an automated irradiation system for production of gaseous radioisotopes applied in industrial processes. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 1583-1592.

    Abstract: The purpose of the present study is to demonstrate how it will be enhanced an Irradiation System (IS) developed with national technology to produce gaseous radioisotopes, by means of the components automation, to avoid the radiation exposure rate to operators of the system, following the ALARA principle (As Low As Reasonably Achievable). Argon-41 (41Ar) and krypton-79 (79Kr) can be produced in continuous scale, gaseous radioisotopes used as radiotracers in industrial process measurements and it can be used in analytical procedures to obtain qualitative and quantitative data systems or in physical and physicochemical studies transfers. The production occurs into the IS, installed in the pool hall of a nuclear research reactor in which the irradiation capsule is positioned near the reactor core containing the isotope gaseous pressurized (40Ar or 78Kr), by (n,γ) reaction and generate the radioisotopes. After the irradiation, the gaseous radioisotope is transferred to the system and, posteriorly, to the storage and transport cylinders, that will be used in an industrial plant. In the first experimental production, was obtained 1.07x1011 Bq (2.9 Ci) of 41Ar distributed in two storage and transport cylinders, operating the IEA-R1 Research Reactor with 4.5 MW and average thermal neutron flux of 4.71x1013 n.cm-2.s-1. However, the system has capacity to five storage and transport cylinders and the estimated maximum activity to be obtained is 7.4x1011 Bq (20 Ci) per irradiation cycle. In this sense, the automation will be based in studies of the production process in the system and the use of Programmable Logic Controllers (PLC), and supervisory software allowing a remote control and consequently better security conditions.

    Palavras-Chave: argon 41; automation; irradiation; krypton 79; neutron flux; production; remote control; thermal neutrons; tracer techniques

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  • IPEN-DOC 26391

    OLIVEIRA, GLAUCIA A.C. de ; LAINETTI, PAULO E.O. ; BUSTILLOS, JOSE O.W.V. ; PIRANI, DEBORA A. ; BERGAMASCHI, VANDERLEI S. ; FERREIRA, JOAO C. ; SENEDA, JOSE A. . Thorium and lithium in Brazil. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5915-5922.

    Abstract: Brazil has one of the largest reserves of thorium in the world, including rare earth minerals. It has developed a great program in the field of nuclear technology for decades, including facilities to produced oxides to microspheres and thorium nitrates. Nowadays, with the current climate change, it is necessary to reduce greenhouse gas emissions, one of this way is exploring the advent of IV Generation reactors, molten salt reactors, that using Thorium and Lithium. Thorium's technology is promising and has been awaiting the return of one nuclear policy that incorporates its relevance to the necessary levels, since countries like the BRICS (without Brazil) have been doing so for years. Brazil has also been developing studies on the purification of lithium, and this one associated to thorium, are the raw material of the molten salt reactors. This paper presents a summary of the thorium and lithium technology that the country already has, and its perspectives to the future.

    Palavras-Chave: lithium; molten salt reactors; nuclear fuels; public policy; purification; thorium; uranium; brazil

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  • IPEN-DOC 26390

    CUNHA, CAIO J.C.M.R.; RODRÍGUEZ, DANIEL G.; LIRA, CARLOS A.B.O.; STEFANI, GIOVANNI L. ; LIMA, FERNANDO R.A.. Thermohydraulic analysis of a fuel element of the AP1000 reactor with the use of mixed oxides of U / Th using the computational fluid dynamic code (CFX). In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5901-5914.

    Abstract: The present work carried out a thermohydraulic analysis of a typical fuel assembly of the reactor AP1000 changing the type of fuel, of UO2 conventionally used for a mixture of oxides of (U,Th)O2 realizing some simplifications in the original design, with the objective to develop of an initial methodology capable of predicting the thermohydraulic behavior of the reactor within the limits established by the manufacturer. An expression for the power density was determined using a coupled neutronic thermohydraulic calculation; once the final expression for power density was determined, the axial and radial temperature profiles in the assembly, as well as the pressure drop and the distribution of the coolant density, were evaluated. Due to the increase in research done on thorium, such as the work of [1], [2], [3], [4] and [5], as well as the mass diffusion of the AP1000, as is the case with [6] and [7]. The present study developed a simplified model, where burnable poisons and spacer grids were not considered, however, it is a consistent model, but with the insertion of these, a more accurate representation of the reactor is expected, providing operational transient analyzes. This tends to strengthen the lines of research that have been carrying out work on the AP1000, as well as in the general sphere of nuclear power plants.

    Palavras-Chave: boundary conditions; burnable poisons; c codes; calculation methods; fuel assemblies; fuel substitution; mixtures; monte carlo method; power density; pwr type reactors; temperature distribution; thermal hydraulics; thorium; transients; uranium dioxide

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  • IPEN-DOC 26389

    SOUZA, PAULA C.A. de; AGUIAR, ANDRE S. ; HEIMLICH, ADINO; LAPA, CELSO M.F.; LAMEGO, FERNANDO. Assessment of potential risk and radiological impact of accidental release from the ARGONAUT reactor. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5877-5885.

    Abstract: In the early days of nuclear energy in Brazil, a reactor designed at the Argonne National Laboratory, originating the name ARGONAUT from the combination of the name of the Laboratory with the initials of Nuclear Assembly for University Training, reached criticality at the Institute of Nuclear Engineering. The Argonaut is a water moderated research reactor, which uses uranium enriched to 20% (235U) with prismatic graphite reflectors, designed to provide a thermal neutron flux up to 1010 n.cm-2.s-1 at an operating power of 5 kW. The presence of a nuclear research facility at the campus of Federal University of Rio de Janeiro (UFRJ) still cause concerns about radiological safety of the community around, even though this facility has been securely operating for more than fifty years. Besides, there were questioning about the potential risk of this facility to the IEN´s workforce by the Central of Harmonization Unit of Brazil (CGU). Thus, the present work aims to assess the potential risk of radiological accidents. Previously, the potential accidents evolving Argonaut reactor were considered to be the insertion of excess reactivity, catastrophic rearrangement of the core, graphite fire and fuel-handling accident. However, a recent accident scenario reassessment concluded that a severe physical damage of the core after reactor shutdown should be the emergency situation with the greater potential risk among the feasible postulated accidents. According with the shutdown procedure, the water, used as moderator and coolant, drains out of the core and the concrete covers (each weighing 2.5 tons) are routinely removed from the top of reactor using a crane. The damage caused by the failure of the crane dropping the covers on the core would lead to breaking of the aluminum coating and the nuclear fuel plates with their release to the reactor room. This study assesses the radiological impact to workers and members of the public caused by partial inventory release to the atmosphere. Generic gaussian model was used to estimate the relative concentrations of air at ground level through the calculation of dispersion factors derived from wind data. For the dose calculation, the conversion coefficients by inhalation and plume immersion established by the ICRP were used. The results show that potential risk is above 1/10 of the limit of annual dose for workers, while they stay below the limit for members of the public, within a radius greater than 1 km.

    Palavras-Chave: argonaut reactor; dose rates; fission product release; fuel elements; gaussian processes; ionizing radiations; personnel; radiation accidents; radiation doses; reactor accidents; risk assessment; volatile matter

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  • IPEN-DOC 26388

    AGUIAR, ANDRE S. ; LEE, SEUNG M. ; SABUNDJIAN, G. . Analysis of the protective actions in the Emergency Planning Zones (EPZs) in the Angra dos Reis region through the calculation of the dose for public individuals due to a severe accident at the Angra 2 Nuclear Plant. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5862-5876.

    Abstract: This work presents the results of the computational simulations of the consequences of a severe accident in Angra 2 nuclear power plant. The severe accident was supposed to be caused by a rupture of 380cm2 in the primary reactor coolant system resulting in loss of coolant. Since the area of the rupture is quite smaller than the total flow area of the pipe of the primary coolant system, 4418cm2, the accident is classified as a small break loss of coolant accident. However, this rupture by itself would not bring the system about a severe accident, which must involve a considerable damage in the nuclear core. Thus, some boundary conditions were added to the problem in order to set a scenario of this kind of accident, which was simulated by means of the MELCOR code. The results obtained by this code show that the release of the radionuclide to the environment starts at the opening of the containment relief valve, and this valve, in turn, opens when the containment pressure reaches 7bar, at 168 hours after the break of the pipe of the coolant system, according to the simulation. The program used for calculation of the release of the radionuclides to the surrounding region of the nuclear plant was the CALMET/CALPUFF code, so that the atmospheric and transport model were elaborated for this code. A source term was used in order to carry out an analysis of the protective actions in the emergency planning zones by means dose calculation for individuals of the public, and it was based on two different scenarios: first scenario, release of the total activity to the atmosphere of Xe, Cs, Ba and Te, after 2h simulation and second scenario, release of the total activity to the atmosphere of Xe, Cs, Ba and Te, after 168h of simulation.

    Palavras-Chave: angra-2 reactor; boundary conditions; c codes; emergency plans; fission product release; loss of coolant; m codes; radiation doses; radiation protection; radioactive materials; radioactivity; reactor accident simulation; severe accidents

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  • IPEN-DOC 26387

    VAZ, ANTONIO C.A. ; RODRIGUES, VALDEMIR G. ; TOYODA, EDUARDO Y. ; SAXENA, RAJENDRA N. . Human factors inclusion proposal in “reactor trip” to increase safety in operation. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5819-5826.

    Abstract: A fundamental concept in nuclear reactor operation is that safety is the result of interactions between human, technological and organizational factors. The National Nuclear Energy Commission understands how human factors from psychological, physiological, behavioral and emotional origin can affect the reactor operation. For that reason, reactor operators are submitted to rigorous evaluations every ye ar. When conducting case study du ring these sixty years of IEA R1, three of them hypothetical and possible related to the reactor operation illustrates the co ncern about the safety and security : Case 1 Operator had a stroke during reactor operation in the control room. C ase 2 Operator suffered stress in traffic in his going to the reactor facility; when performing test in the emergency cooling system for reactor start up, he didn’t close a valve completely; changing the pool water technical quality causing a week delay in the reactor op eration . Case 3 Operator just arrived to afternoon shift in the control room, after a few minutes his co worker noticed that his cognition and behavior has changed, later in the hospital he was diagnosed with head cancer. This interdisciplinary work aims to include human factors of psychological , physiological and behavioral origin in 'reactor trip'. The ‘reactor trip’ (also know n as ‘scram’) usually applies to technical factors to avoid high consequence event, are protection circuits that can assume the s tatus of alert, hazard and essentially shut down the reactor automatically; when temperature, radioactivity, pressure, water flow, voltage and so on ; are out of the operating limits. Technologies associated with neuroscience and psychological assessments s uch as: Face Reader, Analogue Visual Mood Scale and Back Depression Inventory ; allows the evaluation of the operator in the control room. However, problems li ke described in the case study should be minimized. This inter disciplinary theoretical work is based on empirical doctoral thesis in progress.

    Palavras-Chave: control rooms; human factors; iear-1 reactor; radiation protection; reactor accidents; reactor operators; reactor safety; scram; security

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  • IPEN-DOC 26386

    SOBREIRO JUNIOR, ADALBERTO R. ; FREIRE, LUCIANO O. ; ANDRADE, DELVONEI A. de . Proposal for a nuclear power-plant ship decomissioning. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 5793-5804.

    Abstract: The goal of this work is to review decommissioning methods for nuclear propulsion ships throughout of survey on decommissioning experience. Governmental regulation typically dictates cleanup of a decommission site. It is satisfying the stringent regulations that prove to be a primary cost driver for decommissioning and waste disposal. Reactor types and sizes, the number of reactors on an individual plant site, and labor costs are among the main factors affecting costs. Thus, it is so important to develop a good recycling policy after nuclear-power plant ship inactivation. This work found that adequate requirements identification must keep economics always in the center of design. Experience shows, except after major catastrophic accidents, nuclear industry may earn public trust by open dialogue with the population and sound engineering practices, searching for right technical solution and great planning for long time. To achieve this goal, this work proposed the following method: firstly, it presents the characteristics of nuclear-powered submarines. Secondly, an approach concerning the decommissioning process of nuclear-powered submarines adopted by the US Navy, Russian Navy, Royal Navy, French Navy and others which brings the past experience on this field, providing some information on history, architectures and hints of reasons for the success or failures of each project. Finally, this works compared the decommissioning processes of these navies under the perspective of the nuclear regulatory process.

    Palavras-Chave: ships; ship propulsion reactors; decommissioning; government policies; toxic materials; nuclear power; nuclear submarines; waste disposal

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  • IPEN-DOC 26385

    SCURO, NIKOLAS L. ; ANGELO, GABRIEL ; ANGELO, E.; TORRES, WALMIR M. ; UMBEHAUN, PEDRO E. ; ANDRADE, DELVONEI A. de . Preliminary numerical analysis of the flow distribution in the core of a research reactor. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5667-5674.

    Abstract: The thermal-hydraulic safety analysis of research reactors establishes the safety criteria to ensure the integrity of the fuel elements in the reactor core. It assures that all core components are being adequately cooled during operation. It is necessary to know if the average mass flow rate (and their standard deviation) among the fuel assemblies are enough to cool the power generated during operation. Once satisfied such condition, it allows the calculation of the maximum heat flux transferred from fuel assemblies to the coolant, and if the maximum cladding temperatures are below the limits set by the safety criteria. Among the objectives, this study presents a methodology for a preliminary three-dimensional numerical analysis of the flow distribution in the core of the IEA-R1 research reactor, under steady state condition. For this, the ANSYS-CFX® commercial code was used to analyze the flow dynamics in the core, and to visualize the velocity field. It was possible to conclude that a homogeneous flow distribution for all standard fuel assemblies were found, with 2.7% deviation from the average mass flow. What turned out to be negligible and can be assumed that there is a homogeneous distribution in the core. Complex structures were find in the computational domain. Once known the core flow dynamics, it allows future studies to determine whether the heat flux and temperature conditions abbeys thermal-hydraulic safety criteria.

    Palavras-Chave: boundary conditions; c codes; flow models; fuel assemblies; iear-1 reactor; numerical analysis; reactor cores; research reactors; safety; steady-state conditions; thermal hydraulics

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  • IPEN-DOC 26384

    CARVALHO, DANIEL S.M. de; MATTAR NETO, MIGUEL . Assessment of ANSYS LS-DYNA capabilities for analysis of drop tests of nuclear fuel element transportation casks. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5551-5563.

    Abstract: During the transportation of fuel elements, the cask has to provide shielding to protect workers, the public and the environment against the effects of radiation, to prevent an unwanted chain reaction, damage caused by heat and also to provide protection against dispersion of the contents. In order to standardize the design of fuel assembly transportation devices by numerical analysis, a set of dynamic analyzes was conducted to converge in a representative way the phenomena found in the drop tests used in the project qualification. Thus, this paper aims to present and discuss updated recommendations for contacts, material models and general configurations in three benchmarks. These benchmarks represent the phenomena found in numerical simulations of drop trials. Moreover, they are important to obtain an adequate correlation with the lowest possible use of computational resources. From the simulations, it was possible to observe the influence of an analysis carried out in plane strain and another one performed with the complete geometry modeled in scale 1:4 in relation to the computational cost and the precision of the results. A methodology was proposed to calibrate the stiffness and the damping control of the contacts and, mainly, their influence on the behavior of the structure.

    Palavras-Chave: benchmarks; boundary conditions; casks; computerized simulation; damping; finite element method; flexibility; fuel assemblies; fuel elements; mathematical models; nuclear fuels; recommendations; testing; transport

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  • IPEN-DOC 26383

    VIEIRA NETO, ANTONIO S. ; OLIVA, AMAURY M. ; SAUER, MARIA E.L.J. ; HUNOLD, MARCOS C. ; OLIVEIRA, PATRICIA da S.P.de ; ANDREA, VINICIUS . Knowledge base about risk and safety of nuclear facilities to support analysts and decision makers. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5513-5522.

    Abstract: Epistemic uncertainty (uncertainty related to lack of knowledge), often found in the documentation of nuclear facility engineering projects, can affect the decision-making process of managers and analysts on safety and risk issues. This article conceptualizes the nature of the major uncertainties involved in engineering projects and describes a knowledge base developed in order to gather data and information related to the project of an Open-Pool Light-water Research Rector (OPLRR) and whose purpose is to assist professionals who work in the áreas of safety, design, operation, and maintenance of nuclear facilities. In order to reduce the epistemic uncertainties that may rise in the project, the OPLRR knowledge base is designed to contain a set of information that allows identifying and facilitating the forwarding of solutions to address inconsistencies, and/or pending issues that may exist in the project. In this sense, the information and the documents related to the project are organized in a graphical and hierarchical architecture, allowing the knowledge base users to quickly and easily obtain information regarding the systems, processes, equipment, and components of the Project. Besides that, a set of documents containing descriptions, reliability data and some other important information about the systems and components are specially created to the knowledge base and it is crucial to reduce epistemic uncertainties, once it raises the issues and the inconsistencies of the project, as well as it clarifies the interrelations between the systems, the functioning of the equipment, their failures modes and the consequences of their failures, and some other data, which are not originally contained in the documents of the project.

    Palavras-Chave: data covariances; decision making; design; information dissemination; knowledge base; maintenance; nuclear facilities; personnel; pool type reactors; reactor cores; risk assessment; safety

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  • IPEN-DOC 26382

    SANCHEZ, ANDREA ; CARLUCCIO, THIAGO; SABUNDJIAN, GAIANE . The cross sections obtained by the serpent code and formatting the input data for the PARCS code using the GenPMAXS code. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5503-5512.

    Abstract: The Purdue Advanced Reactor Core Simulator (PARCS) is a computer code that solves the time-dependent two-group neutron diffusion equation in three-dimensional Cartesian geometry using nodal methods to obtain the transient neutron flux distribution. The code is used in the analysis of reactivity-initiated accidents in light-water reactors where spatial effects may be important. It may be run in the stand-alone mode or coupled to other NRC thermal-hydraulic codes such as RELAP5. The PARCS neutron code accepts libraries from HELIOS, TRITON, WIMS, SERPENT, etc., codes, but for some libraries is required special formatting. In the case of the SERPENT code, the GenPMAXS code must be used for the PARCS code to be able to read the cross sections library correctly. This work is part of a study on the PARCS/RELAP5 coupling for analyzing the control rod ejection of the Angra 2 reactor core. For this case, the core cross sections were obtained for 6 different branches varying the fuel temperature, moderator temperature, moderator density, boron concentration and considering rods removed and inserted. After obtaining the cross sections with the code SERPENT 2.1.26, these data were passed by a special formatting realized with the code GenPMAXS v6.2. Since GenPMAXS has several options controlling how to process the cross-sections generated by Serpent, a several doubts arose about the correct use of the code. When the doubts are answered, the file with the input data that will be used for the PARCS / RELAP coupling can be built.

    Palavras-Chave: angra-2 reactor; computerized simulation; control elements; coupling; cross sections; monte carlo method; p codes; reactor cores; rod ejection accidents; s codes

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  • IPEN-DOC 26381

    BORGES, EDUARDO M. ; SABUNDJIAN, GAIANE . Small break loss of coolant accident of 200 cm² in cold leg of primary loop of ANGRA 2 nuclear power reactor evaluation. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5479-5490.

    Abstract: The aim of this paper is evaluated the consequences to ANGRA 2 nuclear power reactor and to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 200cm2 of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. The results obtained for ANGRA 2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core.

    Palavras-Chave: angra-2 reactor; cladding; eccs; heat transfer; primary coolant circuits; reactor accident simulation; reactor cores; sbloca; steady-state conditions; two-phase flow; void fraction

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  • IPEN-DOC 26380

    BORGES, EDUARDO M. ; SABUNDJIAN, GAIANE ; BRAZ FILHO, FRANCISCO A.; GUIMARÃES, LAMARTINE N.F.. RELAP5 code simulation of the small break loss of coolant accident of 80 cm² in the cold leg of Angra2 primary loop. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5469-5478.

    Abstract: The aim of this paper was to simulate and evaluate the basic design accident of 80 cm² small break loss of coolant accident (SBLOCA) in the cold leg of the primary loop of the Angra2 nuclear power plant. In this simulation, it was verified that the actuation logics of the Angra2 Reactor Protection System (RPS) and the Emergency Core Cooling System (ECCS) used in this simulation worked correctly, maintaining core integrity with acceptable temperatures throughout the event. The results obtained were satisfactory when compared with those presented by the Angra2 Final Safety Analysis Report (FSAR/A2).

    Palavras-Chave: actuators; angra-2 reactor; boundary conditions; primary coolant circuits; r codes; reactor accident simulation; reactor cooling systems; reactor cores; reactor protection systems; safety analysis; sbloca; steady-state conditions; void fraction

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  • IPEN-DOC 26379

    SOARES, HUMBERTO V.; BELCHIOR JUNIOR, ANTONIO ; FREITAS, ROBERTO L.. RELAP5 modeling of a siphon break effect on the Brazilian Multipurpose Reactor. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5443-5456.

    Abstract: This work presents the thermo-hydraulic simulation of the Brazilian Multipurpose Reactor (RMB) using the RELAP5/Mod3 code. The RMB will provide Brazil with a fundamental infrastructure for the national development on activities of the nuclear sector in the areas of social, strategic, industrial applications and scientific and technological development. A RELAP5/Mod3 code model was developed for thermo-hydraulic simulation of the RMB to analyze the phenomenology of the Siphon Breakers device (four flap valves in the cold leg and one open tube for the atmosphere in the hot leg) during a Loss of Coolant Accident (LOCA) at different points in the primary circuit. The Siphon Breaker device is an important passive safety system for research reactors in order to guarantee the water level in the core under accidental conditions. Different simulations were carried out at different location in the Core Cooling System (CCS) of the RMB, for example: LOCA before the CCS pumps with and without pump trip and LOCA after the CCS pumps and the heat exchanger. In all RELAP5/Mod3 code simulations, the Siphon Breaker device's performance after a LOCA was effective to allow enough air to enter the outlet pipe of the CCS in order to break the siphon effect and preventing the pool level from reaching the riser (chimney) and the RMB core discovering. In all cases, the reactor pool level stabilized at about 5.5 m after the end of the LOCA simulation and the fuel elements were kept underwater and cooled.

    Palavras-Chave: cooling systems; fuel elements; loss of coolant; r codes; reactor accident simulation; reactor cores; reactor safety; rmb reactor; thermal hydraulics; transients

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  • IPEN-DOC 26378

    GOMES, DANIEL de S. ; GIOVEDI, CLAUDIA. Combining probabilistic and deterministic methods for accident analysis. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5429-5442.

    Abstract: This study describes a practical method applied to nuclear reactor safety analysis (NRSA), based on an approach so-called best estimate plus uncertainty (BEPU). The innovative analysis approach involves statistical methods integrated with deterministic rules to fuel licensing code (FLC). The goal of NRSA is to improve safety margins in the nuclear reactor operation, which has partially achieved with uncertainty treatment. Previously, BEPU analysis was widely used to study the loss of coolant accident (LOCA), via inclusion in thermal-hydraulic codes (THC). The systems can measure the impact caused by uncertainties spread in core reactors with a coupling of THC and optimization packages. This paper shows the result of applying the UA/SA technique to FRAPCON, joined with DAKOTA toolkit. This integration will offer the probabilistic analysis coupled with empirical rules. A perfect fusion of the concepts permits the exploration of parametric uncertainties and calibration of physical models. We can use the combined utilization of FLC systems and the DAKOTA toolkit to produce sensitivity analysis. The first step in this approach is to identify all uncertainty sources of the physical models, the reactor design, and manufacturing parameters. It is subsequently used into an FLC, such as FRAPCON, as input parameters. The uncertainties usually distributed using the Wilks formula, which determines the number of samples required for unilateral tolerance. According to Wilks' method, it needs 59 data samples to achieve a confidence level of 95%. Results from Wilks formula found via Monte Carlo simulation, which applies to FLC coupled with sensitivity analysis.

    Palavras-Chave: cladding; data covariances; deterministic estimation; f codes; fuel rods; loss of coolant; probabilistic estimation; reactivity; reactor accidents; reactor cores; reactors; safety analysis; sensitivity analysis; transients

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  • IPEN-DOC 26377

    GOMES, DANIEL de S. ; GIOVEDI, CLAUDIA. Comparative analysis of silicon carbide with zirconium-based alloys. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5417-5428.

    Abstract: According to international plans, the nuclear reactor fleet should reduce operational risk and avoid severe accidents. Around the world, there are 450 nuclear power reactors in operation, which supply about 11% of the electricity consumed. There are programs, such as Advanced Fuels Campaign (AFC), that plan to develop a more tolerant fuel system by 2025. These plans follow security concepts that present two options capable of replacing zirconium alloys used as cladding. The better candidates are metallic alloys and ceramic materials. Until the mid-1970s, austenitic steel was the main coating option. Recently, iron-based alloys have become short-term solutions composed of iron-chromium-aluminum (FeCrAl) alloys. However, there are various advantages from using multilayer of silicon carbide (SIC) and ceramic composites. Silicon carbide has higher corrosion resistance, coupled with higher mechanical strength compared to zirconium alloys. Upon steam contact, ceramic cladding mitigates hydrogen buildup, avoiding explosion risk. This study presents a comparison of the thermal and mechanical properties between zirconium alloys and ceramic alternatives. Ceramic materials show desirable mechanical strength, such as high initial crack resistance, stiffness, ultimate strength, impact response, and high corrosion resistance. SIC has a lower neutron cross-section with significant safety margins.

    Palavras-Chave: ceramics; cladding; comparative evaluations; corrosion protection; cross sections; f codes; fuel rods; mechanical properties; nuclear fuels; physical properties; silicon carbides; steady-state conditions; thermal expansion; zirconium alloys

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  • IPEN-DOC 26376

    GABE, CESAR A.; FREIRE, LUCIANO O. ; ANDRADE, DELVONEI A. de . Modeling dynamic scenarios for safety, reliability, availability and maintainability analysis. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5393-5400.

    Abstract: Safety analysis uses probability combinatorial models like fault tree and/or event tree. Such methods have static basic events and do not consider complex scenarios of dynamic reliability, leading to conservative results. Reliability, availability, and maintainability (RAM) analysis using reliability block diagram (RBD) experience the same limitations. Continuous Markov chains model dynamic reliability scenarios but suffer from other limitations like states explosion and restriction of exponential life distribution only. Markov Regenerative Stochastic Petri Nets oblige complex mathematical formalism and still subject to state explosions for large systems. In the design of complex systems, distinct teams make safety and RAM analyses, each one adopting tools better fitting their own needs. Teams using different tools turns obscure the detection of problems and their correction is even harder. This work aims to improve design quality, reduce design conservatism, and ensure consistency by proposing a single and powerful tool to perform any probabilistic analysis. The suggested tool is the Stochastic Colored class of Petri Nets, which supplies hierarchical organization, a set of options for life distributions, dynamic reliability scenarios and simple and easy construction for large systems. This work also proposes more quality rules to assure model consistency. Such method for probabilistic analysis may have the effect of shifting systems design from “redundancy, segregation and independency” approach to “maintainability, maintenance and contingency procedures” approach. By modeling complex human and automated interventional scenarios, this method reduces capital costs and keeps safety and availability of systems.

    Palavras-Chave: availability; computerized simulation; dynamical systems; maintenance; probabilistic estimation; redundancy; reliability; safety analysis; sensitivity analysis; stochastic processes

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  • IPEN-DOC 26375

    BELCHIOR JUNIOR, ANTONIO ; SOARES, HUMBERTO V.; FREITAS, ROBERTO L.. Validation of the RELAP5 code for the simulation of the Siphon Break effect in pool type research reactors. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5383-5392.

    Abstract: In an open pool type reactor, the pool water inventory should act as a heat sink to provide emergency reactor core cooling. In the Brazilian Multipurpose Reactor – RMB, to avoid the loss of pool water inventory, all the Core Cooling System (CCS) lines penetrate at the pool top, far above the reactor core level. However, as most of CCS equipment and lines are located below the reactor core level, in the case of a Loss of Coolant Accident (LOCA), a large amount of pool water could be lost drained by siphon effect. To avoid RMB research reactor core discovering in the case of a LOCA, siphon breakers, that allow CCS line air intake, are installed in the CCS lines in order to stop the reactor pool draining due to siphon effect. As siphon breakers are important passive safety devices, their effectiveness should be verified. Several previous numerical and experimental studies about siphon break effect were found in the literature. Some of them comment about the effectiveness of the siphon breakers based on their air intake area. Others state that one-dimensional thermo-hydraulic system codes such as RELAP5 code would fail when modeling the siphon break effect. This work shows the RELAP5/MOD3.3 code capability in modeling the siphon break effect. A nodalization for RELAP5/MOD3.3 code of a Siphon Breaker Test Facility located at POSTECH University in Korea was developed. Experiments considering several siphon breakers device intake areas were simulated. A very good agreement between numerical and experimental results was obtained. As siphon breakers intake areas decrease, the siphon breaker effectiveness also decreases and more water is drained from the reactor pool. For smaller siphon breaker intake areas, RELAP5/MOD3.3 code showed conservative results, overestimating the reactor pool water losses.

    Palavras-Chave: computerized simulation; loss of coolant; pipes; pool type reactors; r codes; reactor cores; ruptures; safety analysis; tanks; test facilities; test facilities; validation; void fraction

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  • IPEN-DOC 26374

    OLIVEIRA, ELLISON A. ; OLIVEIRA, PATRICIA S.P. ; MATTAR NETO, MIGUEL ; MATURANA, MARCOS C.. Overview of the seismic probabilistic safety assessment applied to a nuclear installation located in a low seismicity zone. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5368-5382.

    Abstract: Permanent concern on the safety of nuclear installations shall be assured in order to maintain the protection of workers, individuals from the public and the environment. Safety analysis methodologies for both approaches, deterministic and probabilistic, have been developed and updated based on operational experience, investigation of past incidents or accidents, and analysis of postulated initiating events. In general terms, the main objectives of a nuclear safety study are the identification of a comprehensive list of accident initiating events, the evaluation of their impact on the installation and the assessment of the total radiological risk resulting from accidents with off-site releases. Among all initiating events and hazards, there are external hazards that continually challenge the safety of a nuclear facility or its nearby area. In particular, seismic events represent a major contributor to the risk of a nuclear facility. Large levels of ground motion induced by earthquakes may be experienced due to the propagation of mechanical waves on the ground, caused by the displacement of tectonic plates. In this context, a seismic hazard analysis can be carried out in order to predict local acceleration levels with the associated uncertainty distribution, allowing an adequate seismic classification of plant structures, systems and components, including installations located in sites with low seismicity. In order to estimate the risk of a nuclear installation concerning accidents induced by seismic events, a Seismic Probabilistic Safety Assessment (Seismic PSA) shall be performed. In this article, a general description of the Seismic PSA methodology is presented, with emphasis on the supporting studies for this assessment. Finally, this study is under the scope of a master degree project at IPEN – CNEN/SP which intends to apply the methodology described in this article to an experimental nuclear installation containing a PWR reactor designed for naval propulsion to be installed in a low seismicity zone in Brazil.

    Palavras-Chave: earthquakes; nuclear facilities; probabilistic estimation; radiation hazards; radiation protection; risk assessment; safety analysis; seismicity

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  • IPEN-DOC 26373

    LEE, SEUNG M. ; LAPA, NELBIA S.; SABUNDJIAN, GAIANE . MELCOR simulation of a severe accident scenario derived from a small break loca in a typical PWR with passive autocatalytic recombiners. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5346-5359.

    Abstract: This work presents the simulation of a severe accident scenario in a referential model of pressurized water reactor, which came about from a rupture of 20cm2 in a cold leg of a reactor cooling system. The simulation was carried out on the MELCOR code using a model elaborated by the Global Research for Safety – Germany, with the passive autocatalytic recombiners implemented in almost every compartment in the containment. The efficacy and effectiveness of this well-known mitigating measure of severe accident management are demonstrated by means of a comparison with the case previously simulated without this measure using the same model. This referential reactor is important and very useful for the independent analysis of severe accidents in the Brazilian Angra 2 nuclear power plant in virtue of the similarity between both of them, so that after some proper modifications on this referential reactor’s model, it could be applied for the study of severe accidents in the other. In this sense, the result presented in this work is to be taken as an important reference for the severe accident analysis of Angra 2.

    Palavras-Chave: boundary conditions; cladding; loss of coolant; m codes; melt-through; pwr type reactors; radiation protection; reactor accident simulation; reactor cooling systems

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  • IPEN-DOC 26372

    LOBO, RAQUEL de M. ; ANDRADE, ARNALDO H.P. de . Advances in the understanding of the mechanisms of iodine-induced SCC cracking in zirconium alloys. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5339-5345.

    Abstract: In pressurized water reactors (PWR) the fuel rod cladding is the first barrier against the spread of fission products. It is therefore essential to guarantee its use in the reactor. Sometimes the production of electricity requires that certain power plants operate in “network monitoring”. The fuel introduced into nuclear power reactors can then undergo so metimes significant power variations. Following a severe reactor power transient, clad failure can occur through a stress corrosion phenomenon (SCC), under the combined action of mechanical stresses and gaseous fission products generated by the fuel pellets. Among those iodine plays a major role, for it may induce SCC in zircaloy. In the early ages of water cooled reactors (PWRs, BWRs or CANDU), series of similar failures took place following sharp startups. Today power increase rates as well as instantaneous local power levels are limited. Indeed, it is well know that cladding failure by iodine induced stress corrosion cracking (I SCC) may occur under pellet cladding interactions (PCI) conditions during power transients in PWRs. In this paper we review the advances in the understanding of these SCC cracking mechanisms of the fuel rod cladding that would then allow better control of the integrity of the clad during the more severe demands related to the operating conditions of th e PWRs.

    Palavras-Chave: cladding; cleavage; computerized tomography; cracking; fuel rods; iodine; nucleation; pitting corrosion; pwr type reactors; slip; stress corrosion; zircaloy 4

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  • IPEN-DOC 26371

    ANDRADE, ARNALDO H.P. de ; MIRANDA, CARLOS A.J. ; LOBO, RAQUEL de M. . Monitoring of the ductile to brittle transition temperature of reactor pressure vessel steels by means of small specimens. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5322-5338.

    Abstract: Neutron irradiation in nuclear power plants (NPPs) lead to microstructural changes in structural materials which induce a shift of the ductile to brittle transition temperature (DBTT) towards higher temperatures. Monitoring of the DBTT in NPP components receives therefore considerable attention. Small specimen testing techniques are developed for characterizing structural components with a limited amount of materials. One of the most used of these miniature testing is the small punch test (SPT) which is based on disc or square shaped specimens. SPTs may be performed from room to cryogenic temperatures, plotting the absorbed energy until rupture, against the test temperature. A ductile region (high energy) and a brittle region (low energy) with a transition between both zones are usually reported. The transition temperature thus obtained, DBTTSPT, is also related through empirical expressions to the transition temperature obtained in CVN tests, DBTTCVN, or in fracture toughness testing. Linear expressions such as DBTTSPT = α DBTTCVN have been used where α is a material characteristic constant. In all cases, the DBTTSPT temperature is much lower than that obtained in the CVN tests. In this paper, we present a short review of the literature on the determination of the DBTT for nuclear reactors pressure vessels steels by those two techniques analyzing the reason for the difference in their value as mentioned before. In dealing with irradiated materials, is a high priority to limit the exposure of the professional to irradiation. Therefore, the use of miniature specimens receives significant attention in the nuclear community. The high cost of irradiation experiments is a further incentive for using small specimen testing techniques.

    Palavras-Chave: ductile-brittle transitions; embrittlement; fracture properties; irradiation; materials testing; miniaturization; monitoring; reactor vessels; steels

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  • IPEN-DOC 26370

    FAINER, GERSON ; FALOPPA, ALTAIR A. ; OLIVEIRA, CARLOS A. de ; JUNQUEIRA, FERNANDO C. ; FIGUEIREDO, CAROLINA D.R. ; SANTOS, MARCELO M. dos ; CARVALHO, DANIEL S.M. ; MATTAR NETO, MIGUEL . Structural integrity analysis of the heavy water reflector tanks of the IPEN/MB-01 Reactor. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5306-5321.

    Abstract: The IPEN/MB-01 is a zero power research reactor designed and built by IPEN in partnership with the Brazilian Navy. This reactor is located in IPEN and began operating in 1988. IPEN/MB-01 has been used as an experimental facility for studies on neutron parameters of nuclear reactors moderated by light water. In 2016, a project to modify the core structure of IPEN/MB-01 Reactor was initiated. This project aims the replacement of the rod-type fuel structure for a plate-type one. In order to optimize the performance of the experiments, four tanks filled with D2O were installed around the core. This new core will contain fuel elements that are similar to the ones that will be used in the Brazilian Multipurpose Reactor. In this paper, a complete structural integrity analysis of the four heavy water reflector tanks installed in IPEN/MB-01 Reactor is presented. A numerical analysis was performed applying the finite element method, using ANSYS software and considering ASME Code VIII, division 2.

    Palavras-Chave: a codes; finite element method; fuel elements; fuel integrity; heavy water; ipen-mb-1 reactor; numerical analysis; reactor cores; stress analysis; tanks

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  • IPEN-DOC 26369

    FAINER, GERSON ; FALOPPA, ALTAIR A. ; ALMEIDA, JOEDSON T. de ; FIGUEIREDO, CAROLINA D.R. ; CARVALHO, DANIEL S.M. ; MATTAR NETO, MIGUEL . Structural assessment of pressurizer V-102 of the circuit Orquídea. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5290-5305.

    Abstract: The Water Experimental Circuit (CEA) was built in IPEN in eighties and had the aim to perform thermal hydraulic experiments, simulating operational condition of Pressurized Water Reactors and Boiling Water Reactors. The CEA operated until 1984 and since then it was decommissioned. In order to do hydrodynamics tests in MTR fuel type elements of nuclear research reactor, in the years 2015, was conceived an experimental circuit named Orquidea, which shall operate with low pressure and temperature. This paper assess the mechanical and structural suitability of the Pressurizer V-102, that was used in the former Water Experimental Circuit (CEA) aiming reuse this vessel in new the circuit. The methodology applied to evaluate the vessel was based on ASME code, Section VIII, Division 1 & 2.

    Palavras-Chave: a codes; flanges; fuel elements; hydrodynamics; mechanical properties; nozzles; numerical solution; pressurizers; pwr type reactors; reactor vessels; stress analysis; thermal hydraulics

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  • IPEN-DOC 26368

    SANTOS, MARCELO M. dos ; MATTAR NETO, MIGUEL ; MANTECON, JAVIER G. . Preliminar mechanical evaluation of the structure of a nuclear plate-type fuel element. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5276-5289.

    Abstract: The improvement in the efficiency and safety aspects of compact nuclear reactors is directly linked to innovations in fuels and in the geometry of fuel elements (F.E), as is the case of plate-type fuel elements. From the mechanical viewpoint, to ensure that the structure of a fuel element is safe to operate in a compact PWR reactor is important to confirm that it meets the functional design requirements for structures of this type and application, present in ANSI/ANS-57.5-1996 and, also, that the stresses resulting from the loads imposed are less than the permissible mechanical limits for their structural materials, in accordance with ASME III, division 1, subsection NB. In order to develop a methodology of mechanical analysis to verify compliance with the criteria of the cited standards, a numerical model of a plate-type fuel element was developed, taking into consideration the main active loads admitted from the full power operation event belonging to the normal operating condition of a compact PWR type nuclear reactor. The results of the analyses demonstrated that the fuel element designed did not show signs of mechanical failure with respect to the modes of plastic collapse and excess of mechanical deformation.

    Palavras-Chave: a codes; c codes; failures; finite element method; fuel elements; mechanical properties; numerical solution; pwr type reactors; steady-state conditions

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  • IPEN-DOC 26367

    BERRETTA, JOSE R.; LIMA, LEONARDO S.; REIS, REGIS ; AGUIAR, AMANDA A. . PCMI effect study in the fuel rod of a PWR reactor type subjected to power ramps. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5270-5275.

    Abstract: PWR reactor type, when subjected to the power ramp regime, a mechanical interaction between the cladding and the UO2 pellet (PCMI) may occur in the fuel rod. To investigate this phenomenon were used two softwares, the first was a modified fuel performance code to verify the behavior of fuel rod with steel cladding and another to analyze structural mechanical behavior. The fuel performance code results show that there is no contact between the pellet and the cladding in the fuel rod, considering the estimated burning under normal conditions of reactor operation. Thus, it was adopted the hypothesis of the interaction pellet-cladding occurrence, generated by pellet fragmentation and relocation, and power ramp simulation conditions independent of the ramp time. The simulations results show that the fuel rod maintains its integrity under the conditions of the adopted hypothesis.

    Palavras-Chave: c codes; design; finite element method; fuel rods; fuel-cladding interactions; mechanical properties; numerical solution; pwr type reactors; stainless steels; steady-state conditions; stress intensity factors; thermal hydraulics

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  • IPEN-DOC 26366

    FIGUEIREDO, CAROLINA D.R. ; MATTAR NETO, MIGUEL . Recommendations for linearization procedure in pressure Vessel-Nozzle intersections. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5249-5258.

    Abstract: The pressure vessel design is a fundamental step during the construction of new pressurized water reactors (PWRs). In these facilities, several safety requirements are necessary to guarantee the protection of workers, community and environment against the release of radioactive materials. The current version of the ASME Code for vessel pressure presents two types of procedures for structural analysis: Design by Standard and Design by Analysis. The Design by Analysis is a more complex procedure and it requires more rigorous analysis and classification of all types of stresses and loading conditions, in order to incorporate smaller safety coefficients. However, precise rules for achieving the various stress categories have not been implemented in the code. For this reason, this work presents a methodology for the stress linearization in nozzle vessel intersections. The used recommendation is that the line constructed for the linearization should be taken out of transitions elements. So a pressure vessel nozzle intersection was modeling, analyzed and verified then a discussion of how to perform the Code verifications was presented, as well as a mapping of stress. The lines that were constructed in pressure vessel between transition and structural elements in the longitudinal plane (0º) and lines in structural elements in the nozzle in the transversal plane (90º) presents higher stresses.

    Palavras-Chave: a codes; design; finite element method; mesh generation; nozzles; pressure vessels; pwr type reactors; stress analysis

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  • IPEN-DOC 26365

    LEE, SEUNG M. ; YORIYAZ, HELIO ; CABRAL, EDUARDO L.L. . Development of neutron shielding for an inertial electrostatic confinement nuclear fusion device. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5088-5095.

    Abstract: This work aims to develop a suitable neutron shielding for an Inertial Electrostatic Confinement Nuclear Fusion device (IECF). Neutrons are generated in the IECF device as results of nuclear fusion reactions and their detection is fundamental for the development of the IECF device, because experimental data is needed to perform efficiency analysis and model validation. Nevertheless, it is essential to moderate the neutrons down to the thermal state to make it possible to detect those using conventional detectors. Therefore, to properly measure the fast neutron generation rate by the IECF device it is necessary to previously elaborate a detailed neutron transport model between the IECF device and the radiation shielding, where the neutron detector will be located. In this work, a model is elaborated using the Monte Carlo N-Particle Code and is used to design the required radiation shielding for the device. Later, the same model will be used to determine the proportionality factor between the fast neutron generation in the IECF device and the thermal neutron population in the shielding.

    Palavras-Chave: dose equivalents; dose rates; electrostatics; fast neutrons; icf devices; inertial confinement; monte carlo method; neutron flux; neutron transport; shielding; thermal neutrons

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  • IPEN-DOC 26364

    GOMES, DANIEL de S. ; SILVA, ANTONIO T. e . Performance analysis of UO2-SiC fuel under normal conditions. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5056-5069.

    Abstract: This study aims to investigate a fuel mixture of silicon carbide (SiC) and uranium dioxide (UO2) and analyze performance when this fuel applies to light-water reactors (LWRs). Utilization of the licensing code, FRAPCON, with UO2 helped to determine the fuel response under normal conditions initially. High thermal conductivity could permit the use of UO2-10 vol% SiC fuel, showing thermal conductivity values that are far superior to the UO2 alone, exceeding 50% at 900 °C. Ultimately, the formulation should reduce gaseous fission products, avoid fuel cracking, and improve safety margins. SiC has excellent physical properties such as chemical stability, a cross-section with low absorption, irradiation resistance, and a higher melting point. There are some benefits for fuels that use carbon composites such as UO2-carbon nanotube (CNT), and UO2-diamonds. The pellets containing fractions of the carbon limit the amount of fissile U-235 and require additional enrichment to produce the same energy. In the past, there have been various attempts to increase the thermal conductivity of UO2. High conductivity is present in uranium nitride (UN), uranium carbide (UC), and UO2 mixed with beryllium oxide (BeO). The production method of UO2-SiC fuels should include the spark plasma sintering (SPS) technique. Advantages of SPS include a lower manufacturing temperature of 1050°C, better results, and reduced processing time of 30 s. SPS can help produce more tolerant fuels, such as UO2-SiC, UO2-carbon nanotube, and diamond powder dispersion in the UO2 matrix. The thermal conductivity of SiC can decrease substantially under irradiation. UO2-diamond has some drawbacks because of graphitization phenomena.

    Palavras-Chave: f codes; mixtures; nuclear fuels; performance; physical properties; plasma; pwr type reactors; silicon carbides; sintering; thermal conductivity; thermal expansion; uranium dioxide; water cooled reactors

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  • IPEN-DOC 26362

    PIRES, MARINA C. ; MARQUES, JOSE R. de O. ; LEAL NETO, RICARDO M. ; DURAZZO, MICHELANGELO . Study of the manufacturing process of gamma-U7%wtMo dispersion fuel plates. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5024-5035.

    Abstract: The search for new materials for nuclear fuels has been developed over the last 50 years, with the main aim of increasing the fuel efficiency during the operation of the reactors. The need to increase the uranium density in fuels to compensate the reduction of enrichment proposes that the UMo alloy is one of the materials that presents better characteristics to be used as fuel: molybdenum is a material that retains the gamma phase of the uranium in low concentrations, which is the only stable phase of uranium under the irradiation conditions, besides having low thermal neutron absorption. Although more advanced studies already provide information on the interaction between UMo and the Al matrix, we still need to study how this material behaves during all processing steps for fuel fabrication. The present work has the objective of to deepen the technological knowledge about the stages of production of dispersion type nuclear fuel, including the comminution process of the UMo alloy. The alloy pulverization made by the hydriding-grinding-dehydriding technique still reveals a large number of unknowns in the process variables. Knowing some parameters already existent in the literature, it is possible to discuss the behavior of the hydriding process and envision improvements to optimize it as well as make it reproducible. Subsequent manufacturing steps for briquette and rolling were performed according to IPEN's expertise and the results indicate that the UMo alloy is mechanically doable and may prove to be a substitute fuel for the current U3Si2 with a higher uranium density.

    Palavras-Chave: briquets; comminution; dispersion nuclear fuels; fuel plates; fuel substitution; hydridation; molybdenum alloys; production; rolling; uranium alloys

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  • IPEN-DOC 26361

    GOMES, DANIEL de S. ; GIOVEDI, CLAUDIA. Importance of uncertainty modelling for nuclear safety analysis. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5010-5023.

    Abstract: The U.S. Nuclear Regulatory Commission (NRC) reviewed the 10CFR50.46c regulations regard the loss-of-coolant-accident (LOCA), and emergency core cooling system (ECCS). In this planned rulemaking named as 10CFR50.46c. New LOCA criteria included the integration of models used to the hydrogen uptake changes equivalent cladding react (ECR), coupled with peak cladding temperature (PCT). This rule inserts the embrittlement mechanism considering the hydrogen buildup as a pre-transient condition, reducing a loss of operational margin. 10CFR50.46c criteria should combine the effects produced from different fields, such as neutronic analysis, thermal-hydraulic, with fuel performance codes. Besides, it should contemplate Best-Estimate Plus Uncertainty (BEPU) practices. Consequently, increases the challenges to safety analysis because of nuclear power plants run for extended periods than planned initially. In these circumstances, nuclear units need to operate on extended life cycles based on safety margins. With a lifespan of 60 years or more, we reviewed the behavior of the structural material on accident scenarios. This work showed the importance of uncertainties created by physical models such as the fission gas release, thermal conductivity, and loss of ductility caused by hydrides.

    Palavras-Chave: cladding; data covariances; f codes; fuel rods; l codes; lifetime extension; loss of coolant; nuclear fuels; safety analysis; sensitivity analysis; steady-state conditions; thermal hydraulics; transients

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  • IPEN-DOC 26360

    GOMES, DANIEL de S. ; STEFANI, GIOVANNI L. de ; OLIVEIRA, FABIO B.V. de . Analysis of a pressurized power reactor using thorium mixed fuel under regular operation. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4996-5009.

    Abstract: This work discusses a parametric study applied to nuclear power generation based on a mixed fuel formed by the composition of thorium-uranium oxide (Th-U)O2. Also, approached in this study the physical neutrons models of a fuel system composed of ThO2 75 wt% and UO2 25 wt%, with 19.5% enrichment of U-235. The thermodynamic features of the thorium-uranium fuel system compared with the properties of uranium dioxide. Thorium-based fuel operating extended fuel cycles reach of over 80 GWd/MTU in a pressurized water reactor (PWR). Homogenous distribution of thorium-based fuel, used on the reactor core, could reduce Pu-239, once U-233 production capacity dependent on Th-232 replacing U-238 in the fuel matrix. The mixed oxide fuel has a lower buildup of Pu-239, causing the linear heat rate distribution slope to flatten and lowering fuel porosity. The release of gaseous fission products models for (Th-U)O2 could have different diffusion coefficients when compared to uranium oxide models. Besides, resulting in lower thermal gradients than UO2 and a reduction in fuel swelling. This parametric study reviews the aspects of radioactive decay chains of uranium and thorium. It founded the simulation using approved nuclear codes, such as SERPENT for neutron physics calculations and the FRAPCON code, which defines the licensing process. The results show that thoria based fuel has a higher performance than UO2 fuel in regular operation and can improve safety margins.

    Palavras-Chave: comparative evaluations; enthalpy; f codes; mixed oxide fuels; performance; pwr type reactors; s codes; thermal conductivity; thorium; uranium oxides

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  • IPEN-DOC 26359

    GOMES, DANIEL S. ; SILVA, ANTONIO T. e ; OLIVEIRA, FABIO B.V. de ; LARANJO, GIOVANNI S. . Behavior of thorium plutonium fuel on light water reactors. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4984-4995.

    Abstract: Designs using thorium-based fuel are preferred when used in compliance with sustainable energy programs, which should preserve uranium deposits and avoid the buildup of transuranic waste products. This study evaluates a method of converting uranium dioxide (UO2) to thorium-based fuel, with a focus on Th-Pu mixed oxide (Th-MOX). Applications of Th-MOX for light water reactors are possible due to inherent benefits over commercial fuels in terms of neutronic properties. The fuel proposed, (Th-Pu)O2, can be helpful because it would consume a significant fraction of existing plutonium. Aside from the reactor core, the proposed fuel could be useful in existing technology, such as in a pressurized water reactor (PWR). However, licensing codes cannot support Th-MOX fuel without implementing adaptations capable of simulating fuel behavior using the FRAPCON code. The (Th-Pu)O2 fuel should show a plutonium content that produces the same total energy release per fuel rod when using UO2 fuel. Thorium is a fertile material and demands a slightly higher plutonium content when used in Th-MOX. Mixed ceramic oxides show thermodynamic responses that depend on the comprising chemical fractions, and there is little information in databases on irradiation effects. The neutronic analysis is carried out using the SERPENT code to quantify transuranic production and compare this production with the original UO2 fuel assembly. Parameters such as delayed neutron fraction and temperature reactivity coefficient are also determined. Through these analytical methods, the viability and sustainability of the proposed new fuel assembly can be demonstrated in a closed fuel cycle.

    Palavras-Chave: closed fuel cycle; computerized simulation; delayed neutron fraction; f codes; monte carlo method; nuclear fuel conversion; nuclear fuels; plutonium; reactivity coefficients; thermal conductivity; thorium; uranium dioxide; water cooled reactors

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  • IPEN-DOC 26358

    NIELSEN, GUILHERME F. ; MORAIS, NATHANAEL W.S. ; SILVA, SELMA L.; LIMA, NELSON B. de . Crystallographic texture of hot rolled uranium-molybdenum alloys. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4962-4970.

    Abstract: The uranium molybdenum (U-Mo) alloys have potential to be used as low enriched uranium nuclear fuel in research, test and power nuclear reactors. U-Mo alloy with composition between 7 and 10 wt% molybdenum shows excellent body centered cubic phase (γ phase) stabilization and presents a good nuclear fuel testing performance. Hot rolling is commonly utilized to produce parallel fuel plate where it promotes bonding the cladding and the fuel alloy. The mechanical deformation generates crystallographic preferential orientation, the texture, which influences the material properties. This work studied the texture evolution in hot rolled U-Mo alloys. The U7.4Mo and U9.5Mo alloys were melted in a vacuum induction furnace, homogenized at 1000°C for 5 h and then hot rolled at 650°C in three height reductions: 50, 65 and 80%. The as-cast and processed alloys microstructures were characterized by optical and electronic microscopies. The crystalline phases and the texture were evaluated by X-ray diffraction (XRD). The as-cast, homogenized and deformed alloys have γ phase. It was found microstructural differences between the U7.4Mo and U9.5Mo alloys. The homogenized treatment showed effective for microsegregation reduction and were not observed substantial grain size increasing. The deformed uranium molybdenum alloys presented strong γ fiber texture (111) <uvw> and moderated α-fiber texture (hkl) <110>.

    Palavras-Chave: chemical composition; crystal structure; deformation; microstructure; molybdenum alloys; nuclear fuels; optical microscopy; rolling; scanning electron microscopy; texture; uranium alloys; x-ray diffraction

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  • IPEN-DOC 26355

    AGUIAR, AMANDA A. ; ABE, ALFREDO ; GIOVEDI, CLAUDIA. Sensitivity analysis of fuel rod parameters in steady state condition using TRANSURANUS code. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4936-4942.

    Abstract: In 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.

    Palavras-Chave: arkansas-2 reactor; burnup; computerized simulation; fuel rods; monte carlo method; neutron flux; sensitivity analysis; steady-state conditions; t codes

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  • IPEN-DOC 26354

    SILVA, MARCONES C.B. da ; SCHOTT, SANDRO M.C. ; MESQUITA, ROBERTO N. de . Development of a real-time focus estimaton software to be applied in two-phase flow imaging using intelligent processing. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4887-4902.

    Abstract: Image processing has been an increasing research area in the last decades, especially due to crescent technological growth allied with lowering production costs. Many scientific applications have searched for establishment of quality norms associated with possible information obtainment from images. A common need from different applications has been the standardization of focus quality metric. The development of new methods for measuring the focus adjustment in order to obtain image quality metric analysis has enabled more reliable and precise data in many different industry and science sectors. Some examples are industrial equipment parts inspection using computational vision to defects classification. This work presents the initial steps to develop a methodology to estimate focus in real time in two-phase flow experiments inside tube with cylindrical geometry. This methodology is initially based on a software module using artificial intelligence methods to estimate image focus. This module is developed in LabVIEW platform using Fuzzy Logic inference base in different traditional digital focus metrics and integrated with digital cameras to increment precision on focus adjustment during two-phase flow experiments. This method will be calibrated to be used on void fraction estimation through image analysis in the natural circulation loop located at the Nuclear Engineering Center (CEN) do Instituto de Pesquisas Energéticas e Nucleares (IPEN). A set of the initial developed software modules will be presented with their respective functionalities, initial results and experimental focus estimated errors.

    Palavras-Chave: artificial intelligence; defects; focusing; focusing; fuzzy logic; image processing; l codes; m codes; natural convection; quality assurance; real time systems; tubes; two-phase flow; void fraction; brazilian cnen

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  • IPEN-DOC 26353

    PALADINO, PATRICIA A. ; SABUNDJIAN, GAIANE ; CABRAL, EDUARDO L.L. ; JULIÃO, ARTHUR P.. Virtual Reality tools for goods, food and beverage irradiation at IPEN's facilities as a nuclear technology teaching motivation. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4855-4863.

    Abstract: In this research a full-fledged and complete Virtual Reality (VR) environment will be wholly developed and then deployed as a kind of innovative means of widespread divulgation of one topic of nuclear science and nuclear technology most interesting application and its teaching; viz, that related to goods, beverages and mainly food irradiation practices, simulating a virtually guided visit to some of IPEN’s facilities and its already installed and operational scientific equipment, namely, the GAMMACELL irradiator, firstly targeting undergraduate and last year high school students and then, later, the interested general public. In this way, several programs and whole VR platforms, such as Unity, are used as powerful, professional tools for games and videogames development and it is expected that the final product will be made available packaged as an instructive videogame to the community of committed and interested users. Therefore, in doing so, some contemporary reasoned and still debated pedagogical recommendations will be handled and met, hopefully increasing students’ curiosity and good aptitudes towards the disseminated use of nuclear technologies nowadays. It is hoped that perhaps a modest contribution against the many undeserved prejudices and odd misconceptions still remaining nowadays regarding nuclear science development, results and applications, will be abated.

    Palavras-Chave: computerized simulation; education; educational tools; food processing; ionizing radiations; radiators; real time systems; video files; brazilian cnen

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  • IPEN-DOC 26352

    SILVA, LEANDRO G.M. e ; SABUNDJIAN, GAIANE . Virtual visit to nuclear research reactor IEA-R1. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4839-4846.

    Abstract: The aim of this paper is to provide students, educators, and the general public with a virtual tool for learning about the peaceful use of nuclear technology and its importance to humanity. Using new technologies available in the market such as smartphones, software for the development of electronic games, virtual reality glasses, among others, we will virtually reproduce the facilities of the IEA-R1 nuclear research reactor, allowing anyone to perform a virtual and interactive visit to these facilities in a safe and didactic way. The use of virtual reality glasses and applications has been shown to be adequate in relation to the objectives proposed here.

    Palavras-Chave: computer codes; computerized simulation; data visualization; educational tools; iear-1 reactor; mobile phones; real time systems; training

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  • IPEN-DOC 26351

    ALMEIDA, RAFAEL S.P. ; ROCHA, MARCELO S. . Numerical model for calculation of hydraulic transiente and fluid-structure interaction in fluid transport systems. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4731-4742.

    Abstract: In this study the effects of Fluid-structure Interaction during hydraulic transients, more precisely water hammer events, in fluid transport systems are investigated. For this purpose, a numerical model was developed to simulate the effects of Fluid-structure Interaction in a system composed of a reservoir with upstream constant level, a straight pipe and a valve coupled downstream, which can be rigidly fixed or free to move. The transfer of energy from the fluid to the structure associated with pressure waves and their effects, that is, the efforts and displacements generated, is taken into account. The Method of Characteristics is used for solving the hyperbolic partial differential equations system, associated with finite differences and linear interpolations procedures. Three coupling mechanisms are modeled: Friction, Poisson, and junction coupling. The proposed numerical procedure is validated by simulation of a benchmark problem and compared to analytical solutions found in the literature. The results indicated that the model is able to reproduce the main effects Fluid-structure Interaction during hydraulic transients in a pipe conveying fluids. List of symbols A - cross-sectional area, m2 c - classical wave speed, celerity, m/s c˜ - FSI wave speed, celerity, m/s D - inner diameter of pipe, m E - Young modulus of pipe wall, Pa e - pipe wall thickness, m FSI - Fluid-Structure Interaction G - shear modulus of pipe wall material, Pa H - pressure head, m K - fluid bulk modulus, Pa L - length, m MOC - Method of Characteristics P - pressure, Pa R - inner radius of pipe, m T - period, s t - time, s u - pipe displacement, m u̇ - pipe velocity, m/s V - cross-sectional fluid velocity, m/s x - axial coordinate, m g - constant, m/s 𝜇 - Poisson ratio

    Palavras-Chave: benchmarks; computerized simulation; coupling; finite difference method; fluid flow; fluid-structure interactions; friction; hydraulics; nuclear poisons; partial differential equations; pipes; transients; water hammer

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  • IPEN-DOC 26349

    MAPRELIAN, EDUARDO ; BELCHIOR JUNIOR, ANTONIO ; TORRES, WALMIR M. . Lower plenum holes for research reactor core flooding: a proposal to improve the safety in design. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4631-4639.

    Abstract: Modern and high power pool type research reactors generally operate with upward flow in the core. They have a chimney above the core, where the heated fluid is suctioned by the pumps. It passes through the decay tank and is sent to the heat exchangers for the cooling and returns to the core. The pipes inside the reactor pool have passive valves (natural circulation valves) that allow the establishment of natural circulation between the core and the pool for the decay heat removal, when the pumps are inoperative. These valves also have the siphon-breaker function in case of Loss of Coolant Accidents (LOCA), avoiding the pool emptying. In some reactors, these valves are located above the core chimney to facilitate the maintenance. When a LOCA causes a water level below these valves, they loose the natural circulation function. If the water level is the same of the chimney top, the available fluid for the core cooling is only that contained in the chimney and core, and a significant quantity of water in the pool is unavailable for core cooling. To bypass this problem during the reactor design phase, the inclusion of small holes of 10 mm of diameter on the lower plenum lateral side is proposed. These holes will allow a flow path between the pool and the core. Theoretical calculations were performed and analyzed for different drilling configurations: 4, 6 8, and 10 holes. A theoretical analysis of the estimated leakage rate during normal operation and evaporation and replacement rates during a hypothetical LOCA were performed. The calculation results showed that the four configurations analyzed are able to supply the water evaporated from chimney. An experiment is being proposed to validate the theoretical calculations and the considered hypotheses.

    Palavras-Chave: core flooding systems; experimental data; flow rate; holes; leaks; loss of coolant; natural convection; pool type reactors; primary coolant circuits; reactor cores; reactor safety; research reactors; theoretical data; valves

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  • IPEN-DOC 26348

    CASTRO, ALFREDO J.A. de ; CEZARIO, PAULO F.S.. Development of a new test section for the experimental analysis of critical velocity in flat plate fuel element for nuclear research reactor. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4570-4577.

    Abstract: The fuel elements of a MTR type nuclear reactor are mostly composed of aluminum containing the core of uranium sílica (U3Si2) dispersed in an aluminum matrix. These plates have a thickness of the order of millimeters and are much longer in relation to their thickness. This configuration, combined with the need for a flow at high flow rates to ensure the cooling of the fuel element in operation, may create problems of mechanical failure of fuel plate. In the case of critical velocity, excessive permanent deflections of the plates can cause blockage of the flow channel in the reactor core and lead to overheating in the plates. In the first work a test section that simulates a plate-like fuel element with three cooling channels was developed. The dimensions of the test section were based on the dimensions of the Fuel Element of the Brazilian Multipurpose Reactor (RMB). The critical velocity was reached with 14.5 m/s leading to the consequent plastic deformation of the flow channel plates. The signals of extensometers from the test section also showed excitation frequencies due to fluid related phenomena, for example: pressure pulse due to cavitations, fluid resonances, etc. The new test section is being designed to allow internal instrumentation and visualization for a better understanding of the fluid structure coupling. With this new section of test we intend to generate data that allow the assembly of a model that can better simulate the phenomenon of critical velocity for the RMB.

    Palavras-Chave: critical velocity; deformation; experimental data; fuel assemblies; fuel elements; fuel plates; mtr reactor; plates; pressure drop; research reactors; testing

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  • IPEN-DOC 26347

    MOREIRA, PRISCILA G. ; STEFANIAK, IZABELA ; ROCHA, MARCELO S. . Analysis of the thermal conductivity of the aqueous-based TiO2 nanofluid for nuclear applications. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4515-4524.

    Abstract: This work aims to investigate the thermophysical properties of T iO 2 nanofluids in water base experimentally and also comparing results to the literature. Exis ting studies indicate that nanofluids presents increase in thermal conductivity compared to the base fluid which in this study will be water, thus, can be classified as promising fluids for heat transport applications. As the proposal is to use it in nuclea r applications, the survey of experimen tal measurements was performed before and after irradiation in the IPEN installations to verify the effect of ionizing radiation on the properties of nanofluids. Thermal conductivity , viscosity and some visualization of nanopar ticles in SEM were carried out in order to understand the behavior of radiation influence on nanofluids and it properties.

    Palavras-Chave: heat transfer; ionizing radiations; nanofluids; nanoparticles; radiation effects; scanning electron microscopy; thermal conductivity; titanium oxides; viscosity; water

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  • IPEN-DOC 26346

    PRADO, ADELK C. ; ANDRADE, DELVONEI A. ; UMBEHAUN, PEDRO E. ; TORRES, WALMIR M. ; BELCHIOR JUNIOR, ANTONIO ; PENHA, ROSANI M.L. . Status of the development of a fuel assembly decay heat calorimeter for the IEA-R1 nuclear research reactor. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4503-4514.

    Abstract: The heat release due to decay of fission products following a nuclear reactor shutdown is important matter for determining cooling requirements as well as for predicting postulated accident consequences. Accurate evaluation of decay heat can also potentially provide independent data for the cross examination of fuel burnup calculations, which is useful where few resources are available for examination of spent fuel. The evaluation of decay heat from unloaded fuel assemblies of the IEA R1 research reactor was proposed in order to seize that opportunity. With that purpose a special measuring device is under development at the Nuclear and Energy Research Institute (IPEN). Since average heat flux as low as 0.1W/cm2 is expected and since decay heat release must be accurately evaluated, the device design had to overcome the difficulties of measuring small amounts of heat released over a large boundary surface. The design had also to ensure the safe cooling of the fuel assemblies and proper radiological protection for the personnel. In view of the tight constraints, a novel design was adopted. The device features a submersible measurement chamber, which allows all measurement procedures to be performed without removing the fuel assemblies from the reactor pool, and an array of semiconductor thermoelectric modules, which provides highly accurate decay power measurements. The assemblage of the device is currently in progress, the main parts have already been acquired or manufactured and key components passed partial tests. Commissioning and main experiments will be performed up to the end of 2019.

    Palavras-Chave: burnup; calorimeters; decay; fuel assemblies; heat flux; iear-1 reactor; nuclear fuels; radiation protection; reactor cooling systems

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  • IPEN-DOC 26345

    MADEIRA, ALZIRA A.; PEREIRA, LUIZ C.M.; SABUNDJIAN, GAIANE . An Angra 2 LBLOCA simulation model for RELAP5MOD3.3 code with uncertainty analysis. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4476-4502.

    Abstract: This paper describes the activities related to the work planned within Project BRA3.01/12 between CNEN and the European Community, relatively to its Task 2.1 (independent uncertainty quantification and sensitivity analysis utilizing the computational tool SUSA for the calculus related to LOCA simulation for licensing matter). SUSA software has been applied to the reference case, a double-ended LBLOCA in Angra 2, simulated with a RELAP5 code nodalization developed by the thermal hydraulic technicians of CNEN and its research institutes. This original nodalization has been improved for the development of the main objective of Task 2.1. The recommendations that our European counterparts provided on the last workshop, held at CNEN in Rio de Janeiro from January 28th to February 2nd, 2018, have been implemented as far as feasible.

    Palavras-Chave: angra-2 reactor; boundary conditions; cladding; data covariances; lbloca; pressure vessels; r codes; reactor accident simulation; reactor cores; s codes; steady-state conditions

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  • IPEN-DOC 26344

    TORRES, WALMIR M. ; UMBEHAUN, PEDRO E. ; MATTAR NETO, MIGUEL ; BELCHIOR JUNIOR, ANTONIO ; FREITAS, ROBERTO L.. RMB experimental program on the hydrodynamical behavior of fuel assemblies. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4440-4449.

    Abstract: The Brazilian Multipurpose Reactor - RMB is a 30 MW pool type research reactor, that uses Materials Testing Reactor - MTR type fuel assemblies. It has a 5x5 square array core with 23 fuel assemblies and two in-core irradiation positions, operating with upward flow and average velocities nearly 10 m/s in the fuel plates channels. The IEA-R1 is a 5 MW pool type research reactor, which also has a 5x5 square array core with 19 standard fuel assemblies, four control fuel assemblies and a central beryllium irradiation device. It operates with downward flow nearly 1.8 m/s in the channels. In order to verify and provide data and information about the dynamical behavior of fuel assemblies under nominal and critical conditions, the experimental circuit ORQUÍDEA is being designed. This information will be very important for the licensing process of the fuel assembly before its use in the reactor core. This circuit will permits upward and downward flow and dynamical behavior of the fuel assemblies and its parts will be tested and verified. Flow rate, temperature, pressure and differential pressure transducers are the instruments of the circuit. Endurance and critical flow velocity tests will be performed. Dummy fuel assemblies will be used in the tests. It will be instrumented with pressure, strain-gages and flow velocity instruments. The COLIBRI experimental circuit is being designed to make tests that allow the studies of the fluid-structure phenomenology of fuel plates similar to those of the RMB fuel assemblies when subjected to high flow velocities, which can induce pressure differences between the channels formed by the fuel plates. Preliminary structural response studies of the plate’s behavior were performed using a Finite Element Analysis model generated by ANSYS Mechanical. The pressure loadings caused by the fluid flow were calculated using a Computational Fluid Dynamics model created with ANSYS CFX. The fluid-structure interactions will be verified for different channel configurations. In this circuit, vibrations and collapse of the dummy fuel plates will be tested. Experimental data will be compared with CFD (Computational Fluid Dynamics) calculations. This work presents a preliminary design for the ORQUÍDEA and COLIBRI experimental circuits to be built at the Instituto de Pesquisas Energéticas e Nucleares - IPEN of the Comissão Nacional de Energia Nuclear - CNEN.

    Palavras-Chave: comparative evaluations; computerized simulation; critical flow; critical velocity; experimental data; finite element method; flow rate; fuel assemblies; fuel plates; hydrodynamics; pressure range mega pa 10-100; rmb reactor; temperature range 0065-0273 k; temperature range 0400-1000 k

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  • IPEN-DOC 26343

    SILVA, GRACIETE S. de A. e ; MURA, LUIS F.L. ; FUGA, RINALDO ; SANTOS, ADIMIR dos . IPEN/MB-01 reactor experiments with nickel reflectors. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4350-4361.

    Abstract: In the validation and verification processes of calculation methodologies and associated nuclear data libraries, the existence of experiments that can be considered benchmarks is of fundamental importance. For this purpose, a set of experiments with heavy material nuclear reflector was performed in the IPEN/MB-01 reactor using nickel plates properly inserted in the west face of the reactor core. A total of 32 plates around 3 mm thick were used in the experiment. The axial width and length were sufficient to cover the entire active reactor core. For each plate placement step, reactivity measurements were made due to their insertion in the core; as well as of the critical position of the equally removed BC1 and BC2 control rods. It could be observed that the increase of neutron absorption and consequent decrease of neutron moderation dominated the whole physics of the problem when few plates of reflective material were inserted (about 3 plates). Thereafter, neutron reflection became important overcoming neutron absorption; the reactivity increased until it surpassed the situation without plate (excess reactivity zero) obtaining an increase (net gain) of reactivity with the 32 plates inserted (about 295 pcm). Therefore, it was observed that the reflected nucleus became more reactive than the nucleus without reflective material. The theoretical analysis using MCNP-5 and ENDF/B-VII.0 nuclear data library showed the physical aspects of neutron absorption and reflection in the heavy reflector considered; however, it presented a discrepancy when fast neutron reflection dominates the physical phenomenon of neutron transport. In order to verify the impact of other models of thermal scattering of hydrogen in water for the computational simulations of the experiments, three models were considered, besides the one used by the ENDF/B-VII.0 library: ENDF/B-VII.0 scattering law; new evaluation of the S (alpha, beta) for hydrogen bound in water performed in Bariloche Atomic Center, Argentina; and the calculated with new released evaluations for (235)U, (238)U and (16)O.

    Palavras-Chave: absorption; benchmarks; control elements; fast neutrons; fuel rods; ipen-mb-1 reactor; monte carlo method; nickel; nuclear data collections; plates; reactivity; reactor cores; reflection; thermal neutrons; thickness

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  • IPEN-DOC 26342

    STEFANI, GIOVANNI L. de ; GENEZINI, FREDERICO A. ; MOREIRA, JOÃO M. de L.; SANTOS, THIAGO A. dos . Optimization on the core of IEA-R1 research reactor for enhance the radioisotopes production. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4164-4176.

    Abstract: In this work a parametric study was carried out to increase the production of radioisotopes in the IEA-R1 reactor. One of the variables directly proportional to isotope production is the magnitude of the neutron flux in which some material is exposed, so the main objective of this work was to increase neutron flux, especially in the center of the reactor in the beryllium element irradiator (EIBe), within the operational and safety limits of the reactor. The study is initiated by defining a default configuration, in which core of the IEA-R1 reactor is modeled with all fresh fuel assemblies to ensure the reduction of variables that affect the data analysis, then para metric studies were performed evaluating, by comparative analysis of the behavior of the relation of neutron flux versus the fuel for the standard configuration. Therefore, another configuration was tested: the changes in the core of graphite reflecting elements for beryllium, as well as, the result due to reactor core compaction. Parameters such as the fraction of delayed neutrons (Beff) and temperature reactivity coefficient are analyzed to ensure that the configuration has the minimum safety requirements for the reactor safe operation. The results of the study demonstrate a large increase in neutron flux magnitude and in particular in the center of the nucleus in the beryllium irradiating element.

    Palavras-Chave: beryllium; delayed neutrons; fuel assemblies; fuel consumption; iear-1 reactor; isotope production; neutron flux; optimization; parametric analysis; reactivity coefficients; reactor cores; thermal neutrons

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  • IPEN-DOC 26341

    SOUZA, GREGÓRIO; CARLUCCIO, THIAGO; SANCHEZ, PRISCILA; ABE, ALFREDO . Neutron flux intercomparison and ex-core neutron detector optimization in a SMR reactor using MCNP6 code and MAVRIC sequence. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4144-4163.

    Abstract: Ex-core neutron detectors are commonly referred as a detector placed outside the reactor pressure vessel and in a typical SMR design its use is employed to reactor control. Due to its position (far from core) neutron flux calculation for ex-core detector purposes is challenging when using Monte Carlo codes, therefore this work presents an intercomparison between two Monte Carlo codes and also a neutron flux analysis (axially and radially) to better positioning the ex-core neutron detectors. Discrepancies regarding energy treatment can be evaluated as the MAVRIC sequence uses a set of cross sections in a multigroup energy structure while MCNP6 uses continuous energy. In this work, neutron flux intercomparison is mostly focused on variance reduction techniques since these codes presents different approaches, mainly because the MAVRIC sequence uses a hybrid approach combining deterministic and probabilistic methods and MCNP6 code uses traditional variance reduction methods. Some Monte Carlo variables such as figure-of-merit, CPU-time and error distributions maps are evaluated, and neutron flux magnitudes compared. To do so, a typical small modular reactor is modeled with the aid of MCNP6 code and the MAVRIC sequence in two different situations: one being a deep subcritical state with an external neutron source for variance reduction techniques comparison and the other a generic start up procedure (control rods removal) for detector position optimization.

    Palavras-Chave: comparative evaluations; control elements; cross sections; finite difference method; graphite moderated reactors; m codes; monte carlo method; neutron detectors; neutron flux; neutron sources; optimization; reactor cores

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  • IPEN-DOC 26340

    JOÃO, THIAGO G.; SANTOS, DIOGO F. dos ; ROSSI, PEDRO C.R.; SOUZA, GREGORIO S. de ; SANTOS, ADIMIR dos . Monte Carlo modeling of the new plate-type core for the Brazilian IPEN/MB-01 research reactor. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4131-4143.

    Abstract: After 30 years of operation, the IPEN/MB-01 research reactor is about to receive a new plate-type core. This replacement is due to the Brazilian Multipurpose Reactor (RMB) needs, the largest project in nuclear engineering taking place in Brazil. The RMB will be a 30MW open pool-type research reactor, keeping the core in a 5x5 configuration (23 fuel elements, made of U3Si2-Al fuel plates, with 3.7 gU/cm3, 19.75% enriched in U-235 and two extra positions available for materials irradiation). The radioisotopes production, material irradiation, nuclear fuels structural testing and the development of scientific and technological research using neutron beams are the main targets of the RMB enterprise. In this way, in order to verify, experimentally, the calculation methods and data libraries used for the Brazilian Multipurpose Reactor design, reactor cell and mesh structures, control rods effectiveness, isothermal reactivity coefficients and core dynamics due to reactivity insertions, the IPEN/MB-01 new plate-type core is being implemented at the Nuclear and Energy Research Institute (IPEN/CNEN-SP), SP-Brazil. It´s a tank-type research reactor. The core has a 4×5 configuration, with 19 fuel elements (U3Si2-Al, 2.8gU/cm³ and 19.75% enriched in U-235), plus one aluminum block (internal irradiation position). As burnable poison, cadmium wires were used, once they are also employed at the RMB project to control the power density and the excess of reactivity during its operation. The core is reflected by four boxes of heavy water (D2O) and its maximum nominal power is 100W. Thereby, a Monte Carlo modeling was developed using the Monte Carlo N-Particle code (MCNP), along with NJOY, for processing the materials nuclear cross sections. This modeling for the IPEN/MB-01 new plate-type core is presented and some neutronic calculations were also depicted in this paper.

    Palavras-Chave: control elements; cross sections; distribution; fuel plates; ipen-mb-1 reactor; mesh generation; monte carlo method; neutron flux; power density; reactivity; reactor cells; reactor cores; rmb reactor

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  • IPEN-DOC 26339

    HONÓRIO, DANIEL H.; JESUS, MARCELO Z.; PERROTTA, JOSE A. ; MOLNARY, LESLIE de ; AQUINO, AFONSO R. . Licensing aspects of the Brazilian Multipurpose Reactor (RMB). In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4078-4091.

    Abstract: The Brazilian Multipurpose Reactor (RMB) is a project funded by the Brazilian Government by means of the Ministry of Science Technology Innovation and Communication. RMB will be the new Brazilian research reactor, constructed to attend three main purposes: radioisotope production, R&D and material testing. It will be sited 125 km away from S~ao Paulo, strategically, at a Nuclear Compound, where a state owned pole of nuclear technology is located. To construct and operate the RMB facilities, as required by the National Environmental Policy, it is necessary, in addition to the nuclear licensing process of the National Nuclear Energy Commission (CNEN), to conduct all the environmental licensing stages with the Brazilian Environmental Agency (IBAMA). Given this regulatory scenario, based on the standards, guidelines and legal requirements of the IAEA, CNEN, IBAMA and other Brazilian o cial agencies, since 2012, the activities required to comply with the protocol for obtaining initial environmental and construction licenses is being implemented. This paper aims to show a timeline about this process, update the community and register further steps. The RMB entrepreneurs carried out the Environmental Impact Assessment issued the Local report for the radioprotection directory and held three public hearings. Those, among other e orts, resulted on the Local Approval License, which was issued by CNEN Deliberative Commission and on the Initial Environmental License issued by IBAMA. Both of these permits were placed in 2015. Since then some activities for complying with the permit conditions is being performed at the site and properly reported in order to obtain the installation license from the agency.

    Palavras-Chave: environmental impacts; environmental protection; hearings; licensing; limiting values; nuclear facilities; rmb reactor; brazilian organizations

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  • IPEN-DOC 26338

    SILVESTRE, LARISSA J.B. ; SOUSA, EMERSON L. ; SABUNDJIAN, GAIANE . Neuroscience technique applied to the medical diagnostic support system. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4008-4017.

    Abstract: Due to the technological evolution in health, the development of software that helps the doctors in his decisions on the diagnosis of the patient has intensified in recent years. However, adherence by doctors in this regard is still small. The literature shows that doctors form a differentiated group of computer users regarding the acceptance of new technologies. This is justified by the fact that they are generally highly time-pressed, dealing with a wide variety of information and vital decisions. In all professions, the decision-making process is present in most everyday situations and it is important to select the best of them. The Decision Support System (SAD) becomes an ally in this process, especially in the area of health in which the Medical Decision Support Systems (SADM) can contribute to better patient care. It is worth remembering that software to support medical diagnosis may present alternative hypotheses, which will broaden the professional's view on information that he may not be currently associating with. An example of this would be the use of a dermatological software that by capturing the image of a spot on the skin may infer the presence or not of the low, medium and high risk, for example, the SKINVISION software available in the market. Prejudice regarding the use of software that supports the medical decisions may affect directly or indirectly the health care for the population. One of the ways to identify whether or not the medical professional has a prejudice in the use of software in their work practice is through neuroscience techniques applied to the use of Implicit Memory Measurement (Implicit Association Testing -TAI), which does not depend on the participant's conscious attention, and their responses are automatic and spontaneous. The purpose of this work is to use the concepts derived from neuroscience to carry out measures of explicit and implicit memory of medical professors and medical students in order to verify the existence or not of prejudices regarding the use of medical decision support software. This paper presents the results of the pre-test applied to specialists, who are doctors who make use of SADM, and medical students who had the discipline of medical informatics, both groups are from the unit of FAPAC / ITPAC -Porto Nacional -TO. The pre-test was performed in order to verify the internal consistency, that is, if the participants of the chosen association words were understood. For the analysis of the results obtained in this work item, the data were stored in MS Excel® spreadsheets and analyzed with the Statistical Software Statistical Package for Social Sciences (SPSS®), version 23.0, for statistical analysis. SPSS® software was used to calculate Cronbach's alpha, a coefficient in order to measure the internal consistency and reliability of the pre-test of this study (FreeIAT). As a result, the Cronbach's alpha value calculated in the pre-test was 0.838 indicating, thus, good internal consistency.

    Palavras-Chave: computer codes; decision making; diagnosis; medical personnel; neoplasms; uses

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  • IPEN-DOC 26337

    SMITH, RICARDO B. ; SALVETTI, TEREZA C. ; TESSARO, ANA P.G. ; MARUMO, JULIO T. ; VICENTE, ROBERTO . Knowledge management in the decommissioning of nuclear facilities in Brazil. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3997-4007.

    Abstract: In the second half of the twentieth century in Brazil, several nuclear facilities were built for the most varied objectives. The largest number of such facilities is at the Nuclear and Energy Research Institute in São Paulo (IPEN-CNEN/SP). For different reasons, some of these facilities had their projects finalized and were deactivated. Some of the equipment was then dismantled, but the respective nuclear and radioactive material remained isolated in the original sites awaiting the proper decommissioning procedures. The Celeste Project is an example of a facility where the nuclear material has been kept, and is subject to Argentine-Brazilian Agency for Accounting and Control of Nuclear Materials (ABACC) periodic inspections. Because of a number of interests, including financial and/or budgeting situations at the institutions, decades have passed without any further action, and the people who withold information and knowledge about these facilities have already moved away from the area or are in the process of. Therefore, this work proposes an analysis about the knowledge management reflecting on the possible consequences for the decommissioning processes, in case of loss of the knowledge acquired.

    Palavras-Chave: decommissioning; historical aspects; information dissemination; information needs; knowledge management; nuclear facilities; radioactive materials; radioactive waste management; safety; brazilian cnen; brazil

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  • IPEN-DOC 26336

    OLIVEIRA, OTAVIO L. de ; BITELLI, ULYSSES D. . Future challenges for IPEN/MB-01 nuclear research reactor. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3980-3988.

    Abstract: Along the last 30 years, the IPEN/MB-01 research reactor (RR) played a key role in the Brazilian Nuclear Program development. In more than 3,660 sessions it was possible to develop several research experiments, train new operators for the Brazilian nuclear power plants (NPP) and form hundreds of new human resources for nuclear area. Nowadays a new core is under deployment in the facility to prototype the Brazilian Multipurpose Research Reactor (RMB) core project. Several challenges, technical and managerial, are being overcome to fulfill the task, so this paper presents the future challenges for the next 30 years of operation, regarding measures to improve the RR utilization. It is expected to attract more students each year, receive researches from abroad, improve the contact with other RRs around the world to exchange experience in safe operation, maintenance and management system and improve the contacts with Brazilian and Latin America universities. In the same way several experiments are planned to be performed, including those related to the NEA/OECD International Benchmark and those related to the undergraduate and graduate courses.

    Palavras-Chave: personnel; reactor commissioning; reactor cores; research programs; training; brazilian cnen; ipen-mb-1 reactor

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  • IPEN-DOC 26335

    LEOCADIO, MEIRILANE S.; IGAMI, MERY P.Z. ; ANDRADE, DELVONEI A. de . Adherence of the IPEN post-graduation program dissertations to the ABNT norms. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3964-3969.

    Abstract: The process of standardizing or normalizing "something" is a reality in various segments of society, from industry, commerce and even services require Technical Standards to confer a quality standard on any and all goods that are produced. The objective of this research was to verify the adherence of the Dissertations defended in the IPEN/USP Post-Graduation Program to the technical standards of ABNT documentation. We analyzed 85 dissertations made available in the Institutional Repository of the Institute, from 2007 to 2016; we chose to evaluate the adhesion of the Abstract, Literature Review, List of References and Page Formatting by means of a Likert Scale standard form. It was observed that 87% of the Abstracts presented were very adequate to the standards, against 12% that were very inadequate. The Literature Review was very adequate in 51% of the projects, although 27% presented as neither very adequate nor very inadequate (neutral). However, the List of References was inadequate to the norms in 69% of the projects. Finally, in the formatting format it was possible to observe that 56% of the projects were in agreement with the rules presented for paging. In this evaluation it was evidenced that the guide of the Institute has exerted a strong influence on the quality of the assignments, thus guaranteeing greater quality in the physical presentation of the dissertations of the IPEN Program.

    Palavras-Chave: knowledge management; nuclear energy; education; document types; quality control; recommendations; standardization; brazilian cnen

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  • IPEN-DOC 26334

    FREITAS NETO, LUIZ G. ; FREIRE, LUCIANO O. ; SANTOS, ADIMIR dos ; ANDRADE, DELVONEI A. de . Potential advantages of molten salt reactor for merchant ship propulsion. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3878-3888.

    Abstract: Operating costs of merchant ships, related to fuel costs, has led the naval industry to search alternatives to the current technologies of propulsion power. A possibility is to employ nuclear reactors like the Russian KLT-40S, which is a pressurized water reactor (PWR) and has experience on civilian surface vessels. However, space and weight are critical factors in a nuclear propulsion project, in addition to operational safety and costs. This work aims at comparing molten salt reactors (MSR) with PWR for merchant ship propulsion. The present study develops a qualitative analysis on weight, volume, overnight costs, fuel costs and nuclear safety. This work compares the architecture and operational conditions of these two types of reactors. The result is that MSR may produce lower amounts of high-activity nuclear tailings and, if it adopts the 233U-thorium cycle, it may have lower risks of proliferating nuclear weapons. Besides proliferation issues, this 4th generation reactor may have lower weight, occupy less space, and achieve the same levels of safety with less investment. Thus, molten salt regenerative reactors using the 233U-thorium cycle are potential candidates for use in ship propulsion.

    Palavras-Chave: comparative evaluations; cost; molten salt reactors; nuclear fuels; nuclear merchant ships; pwr type reactors; radiation protection; ship propulsion reactors; volume; weight

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  • IPEN-DOC 26333

    D’ERRICO, FRANCESCO; JUNOT, DANILO O. ; POLO, IVON O. ; CHIERICI, ANDREA; CIOLINI, RICCARDO; SOUZA, DIVANIZIA N.; CALDAS, LINDA V. E. ; SOUZA, SUSANA O.. Differential-fading optically stimulable materials for nuclear safeguards. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3838-3843.

    Abstract: Safeguards agencies are concerned with the safety of nuclear installations and the security of nuclear materials. Material protection, control, and accountancy are the first steps towards maintaining continuity of knowledge of these materials and preventing illicit trafficking or diversion of these materials for illicit purposes. Related concerns also exist in arms control, where the item chain of custody is important. In order to strengthen and improve the efficiency and effectiveness of existing safeguards measures, tamperproof devices and materials are needed capable of determining elapsed time since the undeclared movement of a source. Our group developed a new approach for surveillance based on passive, solid-state devices. Relying on a non-electronic detection mechanism is highly desirable because complex electronic components and circuits are potentially vulnerable to tampering and snooping. The device is a set of passive optically stimulated luminescent detectors based on calcium sulfate doped with various rare earths. The different doping produces different temporal fading profiles. When a source causes energy deposition in the detectors, the latter accumulate trapped electrons that undergo de-trapping at different rates. Thus, reading them out produces a set of signals that correlates both with the strength of the source and with the time of its passage.

    Palavras-Chave: calcium sulfates; doped materials; radiation detectors; radioactive materials; rare earths; safeguards; security; thermoluminescent dosemeters

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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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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.

O gerenciamento do Repositório está a cargo da Biblioteca do IPEN. Constam neste RI, até o presente momento 20.950 itens que tanto podem ser artigos de periódicos ou de eventos nacionais e internacionais, dissertações e teses, livros, capítulo de livros e relatórios técnicos. Para participar do RI-IPEN é necessário que pelo menos um dos autores tenha vínculo acadêmico ou funcional com o Instituto. Nesta primeira etapa de funcionamento do RI, a coleta das publicações é realizada periodicamente pela equipe da Biblioteca do IPEN, extraindo os dados das bases internacionais tais como a Web of Science, Scopus, INIS, SciElo além de verificar o Currículo Lattes. O RI-IPEN apresenta também um aspecto inovador no seu funcionamento. Por meio de metadados específicos ele está vinculado ao sistema de gerenciamento das atividades do Plano Diretor anual do IPEN (SIGEPI). Com o objetivo de fornecer dados numéricos para a elaboração dos indicadores da Produção Cientifica Institucional, disponibiliza uma tabela estatística registrando em tempo real a inserção de novos itens. Foi criado um metadado que contém um número único para cada integrante da comunidade científica do IPEN. Esse metadado se transformou em um filtro que ao ser acionado apresenta todos os trabalhos de um determinado autor independente das variáveis na forma de citação do seu nome.