Repositório Digital - IPEN/SP: Recent submissions

  • IPEN-DOC 26586

    MENZEL, SILVIO C. . Proposta de adequação das instalações de infraestrutura – Sala de equipamentos de informática do CPD/SEGRS – Transferência do Rack dos Módulos Clusters do CEN – Instalação elétrica provisória. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Novembro, 2019. (IPEN-CEN-PSE-SEGRS-001-00-RELT-001-00). Restrito.

    Título do projeto: Laboratório Multiusuário de Computação Científica do DIPEN

    Abstract: Apresentar as modificações necessárias para a adequação das instalações de infraestrutura da sala de equipamentos de informática do setor de Gestão de Redes e Suporte Técnico, SEGRS, viabilizando a instalação de um Rack de Módulos Clusters, RMC e respectivos Módulos de Potência Ininterruptos, MPIs (no breaks) e acessórios transferidos do Centro de Engenharia Nuclear, CEN, com o objetivo da implantação do Laboratório Multiusuário de Computação Científica, LMCC, dentro da área do Instituto de Pesquisas Energéticas e Nucleares, IPEN, localizado na Av. Professor Lineu Prestes, 2.242, Portaria Sul, Butantã, na Cidade Universitária "Armando de Salles Oliveira”, CEP 05508-000, Cidade de São Paulo – SP. Este primeiro relatório aborda as modificações necessárias para a alimentação elétrica provisória do Rack de Módulos Clusters, RMC, e o escopo dos serviços deverá abranger apenas a área de elétrica e manutenção de MPIs (No Breaks) existentes. As instalações de ar condicionado, adequações no arranjo da nova sala de MPIs com a montagem de divisórias e a parte de construção civil serão abordadas no segundo relatório para a modificação definitiva da infraestrutura, quando a sala de MPIs receberá 4 (quatro) MPIs novos de 20 KVA e seus gabinetes de baterias externas.

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  • IPEN-DOC 26585

    SHORTO, JULIAN M.B. ; MOLNARY, LESLIE de ; OLIVEIRA, PATRICIA S.P. de ; YAMAGUCHI, MITSUO . Análise qualitativa e quantitativa do Máximo Acidente Hipotético para o Reator IPEN/MB-01. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Setembro, 2019. (IPEN-CEN-PSE-RMB-005-00-RELT-115-00). Restrito.

    Título do projeto: Reator Multipropósito Brasileiro - RMB

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  • IPEN-DOC 26584

    JUNQUEIRA, FERNANDO C. ; SANCHEZ, ANDREA . Manual de utilização do programa X-RAnalysis. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Agosto, 2019. (IPEN-CEN-PSE-RMB-005-00-RELT-102-00). Restrito.

    Título do projeto: Reator Multipropósito Brasileiro - RMB

    Abstract: Este relatório é uma breve descrição da utilização do programa X-RAnalysis para a determinação da densidade superficial de urânio em placas combustíveis do núcleo tipo placa do reator IPEN/MB-01. O programa pode ter seus parâmetros adaptados para placas de urânio em matriz metálica com especificações diferentes das adotadas para o reator IPEN/MB-01. O programa X-RAnalysis foi desenvolvido pelo servidor IPEN Flávio Betti, que se aposentou em 2018 e, este relatório, visa preservar as informações utilizadas por ele durante o trabalho realizado.

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  • IPEN-DOC 26583

    BETTI, FLAVIO . Calibração das radiografias digitais das placas dos elementos combustíveis para o reator IPEN/MB-01. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Setembro, 2019. (IPEN-CEN-PSE-RMB-005-00-RELT-101-00). Restrito.

    Título do projeto: Reator Multipropósito Brasileiro - RMB

    Abstract: Este relatório apresenta os principais aspectos técnicos relacionados ao método proposto para a calibração das radiografias digitais das placas combustíveis produzidas no CCN de acordo com especificações técnicas previamente estabelecidas pelo CEN. Entenda-se como calibração a caracterização dimensional em 2-D do núcleo (cerne) e da moldura, bem como a determinação da concentração (densidade) superficial de urânio expressa, por exemplo, em mg/cm2.

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  • IPEN-DOC 26582

    MENZEL, SILVIO C. . Melhoria do sistema de iluminação - Reator IPEN/MB-01 - Área da Célula Crítica - Especificação de serviços, croquis e materiais - Primeira etapa. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Outubro, 2019. (IPEN-CEN-PSE-RMB-005-00-RELT-098-00). Restrito.

    Título do projeto: Reator Multipropósito Brasileiro - RMB

    Abstract: Este documento apresenta a especificação de serviços, croquis e materiais para a execução da primeira etapa de melhorias nas instalações do sistema de iluminação interna da Área da Célula Crítica do Reator de Pesquisa IPEN/MB-01 do Centro de Engenharia Nuclear, CEN, dentro da área do Instituto de Pesquisas Energéticas e Nucleares, IPEN, localizado na Av. Professor Lineu Prestes, 2.242, Portaria Sul, Butantã, na Cidade Universitária "Armando de Salles Oliveira”, CEP 05508-000, Cidade de São Paulo – SP.

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  • IPEN-DOC 26581

    MOLNARY, LESLIE de . Instrução operacional de meio ambiente para o acompanhamento dos dados meteorológicos. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Maio, 2019. (IPEN-CEN-PSE-INB-006-00-RELT-002-01). Restrito.

    Título do projeto: Comissionamento da Torre Meteorológica – INB Resende

    Abstract: Esse documento apresenta a Revisão 1 do documento IPEN-CEN-PSE-INB-006-00-RELT-002-00, incluindo os comentários e observações do corpo técnico da INB, para a proposta de uma Instrução Operacional de Meio Ambiente (IOMA) para o acompanhamento periódico dos dados meteorológicos coletados em torre instalada no sítio das Indústrias Nucleares do Brasil SA (INB) localizado em Resende (RJ). Essa IOMA deverá ser utilizada pelos analistas e técnicos da Coordenação de Meio Ambiente e Proteção Radiológica Ambiental (COMAP.N) e em situações com interface a outros setores. A IOMA sugere as ações que deverão ser desenvolvidas pela COMAP.N para o acompanhamento dos dados meteorológicos do sítio da INB através de um conjunto de telas de visualização, baseadas no programa LoggerNet (Datalogger Support Software) da Campbell Scientific, Inc. Esse novo conjunto de telas de visualização podem substituir as telas do supervisório atualmente utilizadas na COMAP.N (solução desenvolvida pela área da GERTI.F). As telas de visualização foram desenvolvidas para, através da leitura dos arquivos gerados pelo sistema de aquisição de dados (Resende_15min.dat e Resende_60min.dat) e disponibilizados na rede da Intranet da INB, permitir um melhor acompanhamento dos dados meteorológicos em tempo real, ou para a elaboração de relatórios de funcionamento e disponibilidade do sistema de meteorologia. Essa IOMA também permitirá que ações corretivas possam ser melhor diagnosticadas e as áreas de apoio da INB acionadas para corrigir os problemas com os sensores meteorológicos e com o datalogger. De maneira complementar, essa IOMA faz parte do processo de comissionamento do sistema de meteorologia da INB junto à DRS/CNEN. O IPEN/CNEN-SP vem apoiando a INB como entidade parceira na área de meteorologia ambiental.

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  • IPEN-DOC 26580

    MOLNARY, LESLIE de . Comissionamento do sistema de monitoração meteorológica da INB (Resende, RJ) – análise de dados – período: 08/Abril/2015 a 07/Abril/2017. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Maio, 2019. , Parte 2 (IPEN-CEN-PSE-INB-006-00-RELT-001-01). Restrito.

    Título do projeto: Comissionamento da Torre Meteorológica – INB Resende

    Abstract: Esse documento apresenta a Revisão 1 do documento IPEN-CEN-PSE-INB-006-RELT-001-00, incluindo os comentários e observações da equipe da INB, a respeito da análise dos dados coletados na Torre Meteorológica existente no sítio das Industrias Nucleares do Brasil SA (INB), localizado na Rodovia Presidente Dutra, Km 330, município de Resende (RJ) para o período de 08/04/2015 a 07/04/2017. Essa análise dos dados meteorológicos faz parte do processo de comissionamento do sistema de monitoração meteorológica da INB junto à DRS/CNEN, e da qual o IPEN/CNEN-SP vem apoiando como entidade parceira da INB. São apresentadas os valores estatísticos dos parâmetros coletados, considerações e comentários de eventuais problemas identificados nos dados que estão registrados nos arquivos gerados pelo sistema de aquisição de dados:, Resende_15min.dat e Resende_60min.dat, e disponibilizados na rede da Intranet da INB. De maneira geral, o sistema de monitoração meteorológica vem apresentando melhoras no seu desempenho funcional em relação ao início das mudanças e alterações de equipamentos e sensores realizadas a partir de meados do ano de 2014. Entretanto, ainda são observadas algumas oscilações nos valores registrados, em particular, com os dados de temperatura e diferencial de temperatura. Ao mesmo tempo, o sistema de monitoração meteorológica tem atendido e mantido os requisitos mínimos de disponibilidade e recuperação dos dados meteorológicos, conforme estabelecido na norma CNEN-NE 1.22.

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  • IPEN-DOC 26579

    MOLNARY, LESLIE de . Comissionamento do sistema de monitoração meteorológica da INB (Resende, RJ) – análise de dados – período: 08/Abril/2015 a 07/Abril/2017. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Maio, 2019. , Parte 1 (IPEN-CEN-PSE-INB-006-00-RELT-001-01). Restrito.

    Título do projeto: Comissionamento da Torre Meteorológica – INB Resende

    Abstract: Esse documento apresenta a Revisão 1 do documento IPEN-CEN-PSE-INB-006-RELT-001-00, incluindo os comentários e observações da equipe da INB, a respeito da análise dos dados coletados na Torre Meteorológica existente no sítio das Industrias Nucleares do Brasil SA (INB), localizado na Rodovia Presidente Dutra, Km 330, município de Resende (RJ) para o período de 08/04/2015 a 07/04/2017. Essa análise dos dados meteorológicos faz parte do processo de comissionamento do sistema de monitoração meteorológica da INB junto à DRS/CNEN, e da qual o IPEN/CNEN-SP vem apoiando como entidade parceira da INB. São apresentadas os valores estatísticos dos parâmetros coletados, considerações e comentários de eventuais problemas identificados nos dados que estão registrados nos arquivos gerados pelo sistema de aquisição de dados:, Resende_15min.dat e Resende_60min.dat, e disponibilizados na rede da Intranet da INB. De maneira geral, o sistema de monitoração meteorológica vem apresentando melhoras no seu desempenho funcional em relação ao início das mudanças e alterações de equipamentos e sensores realizadas a partir de meados do ano de 2014. Entretanto, ainda são observadas algumas oscilações nos valores registrados, em particular, com os dados de temperatura e diferencial de temperatura. Ao mesmo tempo, o sistema de monitoração meteorológica tem atendido e mantido os requisitos mínimos de disponibilidade e recuperação dos dados meteorológicos, conforme estabelecido na norma CNEN-NE 1.22.

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  • IPEN-DOC 26578

    MOLNARY, LESLIE de . Relatório semestral de rejeitos e de liberação de efluentes de Angra 1 - dados meteorológicos do 1º semestre de 2018. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Abril, 2019. (IPEN-CEN-PSE-ETN-221-00-RELT-008-00). Restrito.

    Título do projeto: Central Nuclear Almirante Álvaro Alberto - Meteorologia

    Abstract: Esse documento apresenta os dados e parâmetros meteorológicos que estarão disponibilizados no Relatório Semestral de Rejeitos e Liberação de Efluentes de Angra 1 - 1º semestre de 2018, publicado pela área DPR.O da Eletronuclear. Entre as informações disponibilizadas estão:  As tabelas de dados meteorológicos horários para cada uma das liberações de efluentes radioativos gasosos ocorrida durante o 1º semestre de 2018;  As tabelas de distribuição da frequência combinada da direção e velocidade do vento em função da classe de estabilidade atmosférica para liberações pela chaminé de Angra 1; e  Os coeficientes mensais de dispersão atmosférica operacional não deplecionado, deplecionado e o coeficiente de deposição estimados durante o 1º semestre de 2018. Todas as informações são obtidas a partir do banco de dados meteorológicos da CNAAA em Angra dos Reis (RJ), e são de responsabilidade da área ALI.T da Eletronuclear. As informações com as datas e a duração de cada liberação de efluentes gasosos foram fornecidas pela área DPR.O. Observação: As tabelas apresentadas nesse documento seguem a numeração específica do Relatório Semestral de Rejeitos e Liberação de Efluentes de Angra 1.

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  • IPEN-DOC 26577

    MOLNARY, LESLIE de . Relatório semestral de rejeitos e de liberação de efluentes de Angra 2 - dados meteorológicos do 1º semestre de 2018. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Abril, 2019. (IPEN-CEN-PSE-ETN-221-00-RELT-007-00). Restrito.

    Título do projeto: Central Nuclear Almirante Álvaro Alberto - Meteorologia

    Abstract: Esse documento apresenta os dados meteorológicos que estarão disponibilizados no Relatório Semestral de Rejeitos e Liberação de Efluentes de Angra 2 - 1º semestre de 2018, publicado pela área DPR.O da Eletronuclear. Entre os dados e informações disponibilizadas estão:  As tabelas de distribuição da frequência combinada da direção e velocidade do vento em função da classe de estabilidade atmosférica para liberações pela chaminé de Angra 2; e  Os coeficientes mensais de dispersão atmosférica operacional não deplecionado, deplecionado, e o coeficiente de deposição estimados durante o 1º semestre de 2018. Todas as informações são obtidas a partir do banco de dados meteorológicos da CNAAA em Angra dos reis (RJ), e são de responsabilidade da área ALI.T da Eletronuclear. Observação: As tabelas apresentadas nesse documento seguem a numeração específica do Relatório Semestral de Rejeitos e Liberação de Efluentes de Angra 2.

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  • IPEN-DOC 26576

    MOLNARY, LESLIE de . Análise dos dados meteorológicos da CNAAA - período Abril de 2019. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Maio, 2019. (IPEN-CEN-PSE-ETN-221-00-RELT-006-00). Restrito.

    Título do projeto: Central Nuclear Almirante Álvaro Alberto - Meteorologia

    Abstract: Esse documento apresenta a análise dos dados meteorológicos de temperatura, precipitação pluviométrica, velocidade e direção do vento no nível de 10 m e a classe de estabilidade atmosférica de Pasquill coletados na Torre A e a velocidade e direção do vento no nível de 15 m da Torre C para o mês de abril de 2019. Esses dados são obtidos através do sistema de meteorologia da Central Nuclear Almirante Álvaro Alberto, no município de Angra dos Reis (RJ).

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  • IPEN-DOC 26575

    MOLNARY, LESLIE de . Dados de precipitação pluviométrica na CNAAA - período Janeiro a Dezembro de 2018. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Maio, 2019. (IPEN-CEN-PSE-ETN-221-00-RELT-005-00). Restrito.

    Título do projeto: Central Nuclear Almirante Álvaro Alberto - Meteorologia

    Abstract: Esse documento apresenta a distribuição dos totais mensais e anual da precipitação pluviométrica observado no decorrer do ano de 2018 no sítio da Central Nuclear Almirante Álvaro Alberto, no município de Angra dos Reis (RJ). A precipitação pluviométrica acumulada em 2018 foi de 2.905,5 mm. O mês mais chuvoso foi Fevereiro/2018 com um total mensal de 477,5 mm. A precipitação máxima de 24 horas (período contínuo) foi de 119,25 mm entre 00 h do dia 14/02 e 24 h do dia 14/02 de2018. Os dados meteorológicos apresentados serão utilizados pela Gerência de Engenharia Civil e Estruturas Metálicas, para compor documento para o estudo da estabilidade das encostas da Serra do Mar na área próxima ao sítio da CNAAA.

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  • IPEN-DOC 26574

    MOLNARY, LESLIE de . Relatório semestral de rejeitos e de liberação de efluentes de Angra 1 - dados meteorológicos do 2º semestre de 2018. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Abril, 2019. (IPEN-CEN-PSE-ETN-221-00-RELT-004-00). Restrito.

    Título do projeto: Central Nuclear Almirante Álvaro Alberto - Meteorologia

    Abstract: Esse documento apresenta os dados e parâmetros meteorológicos que estarão disponibilizados no Relatório Semestral de Rejeitos e Liberação de Efluentes de Angra 1 - 2º semestre de 2018, publicado pela área DPR.O da Eletronuclear. Entre as informações disponibilizadas estão:  As tabelas de dados meteorológicos horários para cada uma das liberações de efluentes radioativos gasosos ocorrida durante o 2º semestre de 2018;  As tabelas de distribuição da frequência combinada da direção e velocidade do vento em função da classe de estabilidade atmosférica para liberações pela chaminé de Angra 1; e  Os coeficientes mensais de dispersão atmosférica operacional não deplecionado, deplecionado e o coeficiente de deposição estimados durante o 2º semestre de 2018. Todas as informações são obtidas a partir do banco de dados meteorológicos da CNAAA em Angra dos Reis (RJ), e são de responsabilidade da áreaALI.T da Eletronuclear. As informações com as datas e a duração de cada liberação de efluentes gasosos foram fornecidas pela área DPR.O. Observação: As tabelas apresentadas nesse documento seguem a numeração específica do Relatório Semestral de Rejeitos e Liberação de Efluentes de Angra 1.

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  • IPEN-DOC 26573

    MOLNARY, LESLIE de . Relatório semestral de rejeitos e de liberação de efluentes de Angra 2 - dados meteorológicos do 2º semestre de 2018. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Abril, 2019. (IPEN-CEN-PSE-ETN-221-00-RELT-003-00). Restrito.

    Título do projeto: Central Nuclear Almirante Álvaro Alberto - Meteorologia

    Abstract: Esse documento apresenta os dados meteorológicos que estarão disponibilizados no Relatório Semestral de Rejeitos e Liberação de Efluentes de Angra 2 - 2º semestre de 2018, publicado pela área DPR.O da Eletronuclear. Entre os dados e informações disponibilizadas estão:  As tabelas de distribuição da frequência combinada da direção e velocidade do vento em função da classe de estabilidade atmosférica para liberações pela chaminé de Angra 2; e  Os coeficientes mensais de dispersão atmosférica operacional não deplecionado, deplecionado, e o coeficiente de deposição estimados durante o 2º semestre de 2018. Todas as informações são obtidas a partir do banco de dados meteorológicos da CNAAA em Angra dos reis (RJ), e são de responsabilidade da área ALI.T da Eletronuclear. Observação: As tabelas apresentadas nesse documento seguem a numeração específica do Relatório Semestral de Rejeitos e Liberação de Efluentes de Angra 2.

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  • IPEN-DOC 26572

    MOLNARY, LESLIE de . Seção 2.3 – Meteorologia – Rev. 1A - Relatório preliminar de análise de segurança da Unidade de Armazenamento a Seco (UAS) - Central Nuclear Almirante Álvaro Alberto. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Março, 2019. (IPEN-CEN-PSE-ETN-221-00-RELT-002-01). Restrito.

    Título do projeto: Central Nuclear Almirante Álvaro Alberto - Meteorologia

    Abstract: Esse documento apresenta o texto da Seção 2.3 – Meteorologia – Revisão 1B - do Relatório Preliminar de Análise de Segurança da Unidade de Armazenamento a Seco (UAS) da Central Nuclear Almirante Álvaro Alberto (CNAAA) localizado no município de Angra dos Reis, Estado do Rio de Janeiro, Brasil. A Revisão 1 procurou atender aos comentários de melhoria de texto, identificação de referências e adequação de terminologia, assim como, responder as exigências formuladas no Parecer Técnico PT-CODRE- 122/18 (SEI no. 0001124) emitido em 26/12/2018 e que faz parte do Ofício 7/2019-CGRC/DRS/CNEN de 21/01/2019. Entretanto, após a entrega do documento IPEN-CEN-PSE-ETN-221-00 - RELT-002-00, a Gerência Técnica da UAS informou que não são previstas liberações atmosféricas de rotina ou em condições de acidente. Assim sendo, a subseção 2.3.4 Atmospheric Dispersion Estimates foi reescrita, eliminado-se a necessidade de apresentar os coeficientes de dispersão de rotina (longa duração) e de acidentes (curta duração). Foi inserida uma tabela descrevendo os diversos cenários possíveis com a UAS para subsidiar a não apresentação dos respectivos coeficientes de dispersão atmosférica. O novo sistema de meteorologia da CNAAA está em fase de implantação e comissionamento e, portanto, os dados meteorológicos coletados ainda não fazem parte do banco de dados disponíveis para avaliação das variáveis meteorológicas locais.

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  • IPEN-DOC 26571

    MOLNARY, LESLIE de . Seção 2.3 – Meteorologia – Rev. 1 - Relatório preliminar de análise de segurança da Unidade de Armazenamento a Seco (UAS) - Central Nuclear Almirante Álvaro Alberto. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Fevereiro, 2019. (IPEN-CEN-PSE-ETN-221-00-RELT-002-00). Restrito.

    Título do projeto: Central Nuclear Almirante Álvaro Alberto - Meteorologia

    Abstract: Esse documento apresenta o texto da Seção 2.3 – Meteorologia – Revisão 1 - do Relatório Preliminar de Análise de Segurança da Unidade de Armazenamento a Seco (UAS) da Central Nuclear Almirante Álvaro Alberto (CNAAA) localizado no município de Angra dos Reis, Estado do Rio de Janeiro, Brasil. A Revisão 1 procurar atender aos comentários de melhoria de texto, identificação de referências e adequação de terminologia, assim como, responder as exigências formuladas no Parecer Técnico PT-CODRE-122/18 (SEI no. 0001124) emitido em 26/12/2018 e que faz parte do Ofício 7/2019-CGRC/DRS/CNEN de 21/01/2019. O novo sistema de meteorologia da CNAAA está em fase de implantação e comissionamento e, portanto, os dados meteorológicos coletados ainda não fazem parte do banco de dados disponíveis para avaliação das variáveis meteorológicas locais.

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  • IPEN-DOC 26570

    MOLNARY, LESLIE de . Análise dos Dados de Precipitação Pluviométrica para os dias 25 a 27 de Novembro de 2018 na CNAAA. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Janeiro, 2019. (IPEN-CEN-PSE-ETN-221-00-RELT-001-00). Restrito.

    Título do projeto: Central Nuclear Almirante Álvaro Alberto - Meteorologia

    Abstract: Esse documento apresenta uma análise dos dados de precipitação pluviométrica no período de 25 a 27 de novembro de 2018 observados na CNAAA. Durante esse dias foram observadas intensidades de precipitação acumuladas em 24 horas de até 96,0 mm. Valores significativos de precipitação, mas não muito superiores aos valores das normais climatológicas observados na estação do INMET/Angra dos Reis para o mês de Novembro (108,4 mm no período 1981- 2010). Essa análise foi solicitada pelo Departamento de Engenharia Civil da Eletronuclear.

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  • IPEN-DOC 26569

    MENZEL, SILVIO C. . Especificação de serviços e materiais – adequação das instalações elétricas – Prédio/Laboratório Van Der Graaf – Bancada do dispositivo de fusão nuclear – CRPQ – Projeto Básico. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Outubro, 2019. , Parte 3 (IPEN-CEN-P&D-CEN-039-00-RELT-002-00). Restrito.

    Título do projeto: Reator de Fusão Nuclear com Confinamento Eletrostático Assistido por Campo Magnético

    Abstract: Apresentar as modificações necessárias para adequação das instalações elétricas de alimentação e distribuição do Prédio/Laboratório Van Der Graaf para a implantação da Bancada do Dispositivo de Fusão Nuclear em área específica do Centro do Reator de Pesquisa, CRPQ, dentro da área do Instituto de Pesquisas Energéticas e Nucleares, IPEN, localizado na Av. Professor Lineu Prestes, 2.242, Portaria Sul, Butantã, na Cidade Universitária "Armando de Salles Oliveira”, CEP 05508- 000, Cidade de São Paulo – SP.

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  • IPEN-DOC 26568

    MENZEL, SILVIO C. . Especificação de serviços e materiais – adequação das instalações elétricas – Prédio/Laboratório Van Der Graaf – Bancada do dispositivo de fusão nuclear – CRPQ – Projeto Básico. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Outubro, 2019. , Parte 2 (IPEN-CEN-P&D-CEN-039-00-RELT-002-00). Restrito.

    Título do projeto: Reator de Fusão Nuclear com Confinamento Eletrostático Assistido por Campo Magnético

    Abstract: Apresentar as modificações necessárias para adequação das instalações elétricas de alimentação e distribuição do Prédio/Laboratório Van Der Graaf para a implantação da Bancada do Dispositivo de Fusão Nuclear em área específica do Centro do Reator de Pesquisa, CRPQ, dentro da área do Instituto de Pesquisas Energéticas e Nucleares, IPEN, localizado na Av. Professor Lineu Prestes, 2.242, Portaria Sul, Butantã, na Cidade Universitária "Armando de Salles Oliveira”, CEP 05508- 000, Cidade de São Paulo – SP.

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  • IPEN-DOC 26567

    MENZEL, SILVIO C. . Especificação de serviços e materiais – adequação das instalações elétricas – Prédio/Laboratório Van Der Graaf – Bancada do dispositivo de fusão nuclear – CRPQ – Projeto Básico. São Paulo: Instituto de Pesquisas Energéticas e Nucleares - CEENG, Outubro, 2019. , Parte 1 (IPEN-CEN-P&D-CEN-039-00-RELT-002-00). Restrito.

    Título do projeto: Reator de Fusão Nuclear com Confinamento Eletrostático Assistido por Campo Magnético

    Abstract: Apresentar as modificações necessárias para adequação das instalações elétricas de alimentação e distribuição do Prédio/Laboratório Van Der Graaf para a implantação da Bancada do Dispositivo de Fusão Nuclear em área específica do Centro do Reator de Pesquisa, CRPQ, dentro da área do Instituto de Pesquisas Energéticas e Nucleares, IPEN, localizado na Av. Professor Lineu Prestes, 2.242, Portaria Sul, Butantã, na Cidade Universitária "Armando de Salles Oliveira”, CEP 05508- 000, Cidade de São Paulo – SP.

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  • IPEN-DOC 26629

    YOSHIMURA, TANIA M. . Luz de baixa potência como proposta terapêutica à síndrome metabólica em modelo animal / Low level light therapy as a therapeutic proposal for mice with metabolic syndrome . 2014. Dissertação (Mestrado em Tecnologia Nuclear) - Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP, São Paulo. 67 p. Orientador: Martha Simões Ribeiro. DOI: 10.11606/D.85.2014.tde-09042015-143109

    Abstract: A síndrome metabólica (SM) é uma condição clínica que agrupa uma variedade de morbidades, como hiperglicemia, pressão arterial elevada, dislipidemia aterogênica e obesidade (particularmente na região abdominal). Nessa conjuntura, os principais tecidos-alvo da ação da insulina sofrem alterações metabólicas que aumentam o risco de ocorrência de doenças cardiovasculares e diabetes tipo 2. As alterações teciduais observadas são caracterizadas por infiltrados de células do sistema imune, especialmente macrófagos. Citocinas pró-inflamatórias, como TNF-α, são liberadas e alcançam a corrente sanguínea, promovendo nesses indivíduos um estado de inflamação crônica e sistêmica. O tecido adiposo intra-abdominal parece ser de particular importância no estabelecimento desse quadro inflamatório, e estratégias direcionadas no sentido de modular os processos inflamatórios nesse tecido podem atenuar as consequências da SM. Os reconhecidos benefícios da terapia com luz de baixa potência em condições inflamatórias nos permitem supor que essa poderia ser uma proposta terapêutica para a SM. Sendo esse o nosso foco de estudo, camundongos adultos, machos, das linhagens C57BL/6 e BALB/c receberam dieta hiperlipídica durante 8 semanas para indução do quadro de SM. Os animais foram então irradiados sobre a superfície abdominal no decorrer de 21 dias, usando um LED (λ = 850 nm, 6 sessões, 300 s por sessão, potência = 60 mW, fluência = 6 J/cm², taxa de fluência = 19 mW/cm²). Antes e durante o tratamento, amostras de sague foram coletadas para quantificação de glicose, colesterol total e triglicérides plasmáticos. Considerando os parâmetros de irradiação adotados, a terapia com luz de baixa potência não se mostrou efetiva para alterar massa corporal, glicemia, colesterol total e triglicérides de camundongos alimentados com dieta hiperlipídica.

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  • IPEN-DOC 26630

    GOMES, ANTONIO M.S. . Determinação da capacidade de adsorção de Cu, Mn e V em biomassa seca de macrófitas / Determination of Cu, Mn and V adsorption capacity in dry macrophyte biomass . 2019. Dissertação (Mestrado em Tecnologia Nuclear) - Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP, São Paulo. 98 p. Orientador: Paulo Sergio Cardoso da Silva. DOI: 10.11606/D.85.2020.tde-27122019-132214

    Abstract: Os corpos hídricos sempre foram expostos a todos os tipos de contaminação, sejam eles naturais ou antropogênicos, desde erupções vulcânicas até a liberação deliberada de esgoto químico sem qualquer tratamento. Nesse contexto, as macrófitas possuem grande importância para os ambientes aquáticos, embora ainda desconhecida para muitos. Dada a importância de pesquisas relacionadas as formas de minimizar os efeitos nocivos causados pela descarga de poluentes, o objetivo deste trabalho foi determinar a capacidade de adsorção, de elementos em soluções aquosas, por biomassa seca obtida de macrófitas (Eichhornia crasssipes, Egeria densa, Pistia stratiotes e Salvinia auriculata), como forma de contribuir para a remoção de contaminantes de efluentes, e também, como forma de utilização da matéria orgânica produzida por estas plantas. Os elementos analisados foram o cobre, manganês e vanádio, este último com poucos trabalhos relatados em literatura. A metodologia utilizada consistiu no cultivo de plantas em um ambiente livre de cargas poluidoras, produção da biomassa, determinação de características físico-químicas, determinação da capacidade de adsorção em função da variação do pH, tempo de contato e concentração dos elementos de interesse na solução. Os métodos analíticos empregados foram análise por ativação neutrônica e espectrometria de absorção atômica com forno de grafite. Os resultados indicaram que as biomassas secas produzidas não adsorveram o cobre. A biomassa de S. auriculata foi a que apresentou a maior capacidade de remoção de Mn e E. Crassipes foi a que apresentou a maior capacidade de remoção de V em soluções aquosas, nas condições em que foram realizados os procedimentos neste trabalho.

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  • IPEN-DOC 26628

    GROSCHE, LUCAS C. . Síntese de material de valor agregado a partir de coproduto da combustão de carvão : caracterização e aplicação na remediação de efluente aquoso / Synthesis of value-added material from coal combustion co-product: characterization and application in aqueous effluent remediation . 2019. Tese (Doutorado em Tecnologia Nuclear) - Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP, São Paulo. 95 p. Orientador: Denise Alves Fungaro. DOI: 10.11606/T.85.2020.tde-05122019-111420

    Abstract: O presente projeto se insere na sugestão encontrada no "Roadmap tecnológico para a produção e uso limpo do carvão mineral nacional: 2012 a 2035" quanto às ações necessárias para que seja estabelecido um ambiente favorável ao maior uso do carvão mineral no Brasil. Isso se dá no que tange ao desenvolvimento de tecnologia considerada prioritária para o setor da geração termelétrica, a saber: aproveitamento de coprodutos da combustão de carvão (PCC). Neste contexto, o projeto envolveu o desenvolvimento de processos capazes de sintetizar material de valor agregado a partir de PCC, como fonte alternativa de silício e alumínio. O material escolhido, proveniente da queima do carvão foi o resíduo de dessulfurização de gases de exaustão, e os nanomateriais sintetizados são considerados produtos de alto valor agregado por possibilitar inúmeras aplicações. O presente trabalho foi dividido em duas etapas; a primeira etapa tratou da síntese e caracterização dos nanomateriais obtidos a partir de diferentes condições da reação de ativação alcalina buscando aperfeiçoar o processo de síntese. Os resíduos de dessulfurização foram coletados de três localidades diferentes, que no início do projeto representavam todas as localidades onde a tecnologia de dessulfurização já estava sendo aplicada na geração de energia elétrica a partir do carvão. Quanto a caracterização dos resíduos, embora existam diferenças envolvidas nos processos que originam as amostras gerando muitas vezes materiais com formas cristalinas diferentes, foram encontrados em todas as amostras os elementos cálcio, alumínio e sílicio. As amostras das três localidades de amostragem foram submetidas ao processo de ativação hidrotérmica alcalina formando principalmente materiais zeolíticos (Sodalitas), tobermoritas e outros compostos do tipo hidrotalcitas. A segunda etapa do trabalho foi direcionada ao uso dos materiais no tratamento de água contaminada com césio, com este objetivo o resíduo de dessulfurização que indicou maior presença de tobermoritas e hidrotalcitas foi selecionado para otimização do processo de síntese hidrotérmica por duas etapas incluindo fusão previa do resíduo. E o material com os melhores resultados de acordo com os materiais cristalinos obtidos foi testado quanto a sua capacidade de remoção de íons de césio em solução, assim como sua seletividade em relação ao sódio presente em certos meios como, por exemplo, na água do mar, em soluções salinas sintéticas e também em uma amostra de água marinha. A capacidade de adsorção do Cs+ sobre o material adsorvente foi de 1949 μmol g-1, indicando que o adsorvente sintetizado pertence ao grupo de materiais com alta capacidade de adsorção de césio quando comparado com outros materiais estudados na literatura, além de apresentar seletividade para o íon de césio em relação à água do mar. Por fim entende-se que o material tem grande potencial para aplicações em remediações em acidentes como o de Fukushima, onde césio radioativo foi liberado na água do mar. Especialmente devido ao material como o resíduo de dessulfurização ser produzido em larga escala e não possuir aplicação na indústria sendo destinado a aterros onde pode se tronar passivo ambiental.

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  • IPEN-DOC 26627

    CAMARGO, ELAINE F. de . Síntese de suportes de eletrocatalisadores para aplicação em células a combustível poliméricas alimentadas por álcoois : óxido de índio dopado com estanho (ITO), óxido de estanho dopado com antimônio (ATO) e óxido de grafeno reduzido (rGO) / Synthesis of electrocatalyst support for application in alcohol-fueled polymer fuel cells: tin doped indium oxide (ITO), antimony doped oxide (ATO) and reduced graphene oxide (rGO) . 2019. Tese (Doutorado em Tecnologia Nuclear) - Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP, São Paulo. 116 p. Orientador: Dolores Ribeiro Ricci Lazar. Coorientador: Almir Oliveira Neto. DOI: 10.11606/T.85.2020.tde-13122019-112147

    Abstract: No sentido de aumentar a eficiência das células a combustível de baixa temperatura de operação, utilizando o etanol como combustível, busca-se desenvolver eletrocatalisadores de alta atividade. Neste estudo, eletrocatalisadores de platina, suportados sobre óxido de índio dopado com estanho (ITO) e de óxido de estanho dopado com antimônio (ATO) foram sintetizados com o intuito de promover a maior eficiência das reações de eletrooxidação do etanol, por promoverem a oxidação do CO intermediário nas reações de oxidação do combustível. A presença de rGO no compósito também foi avaliada, considerando que os grupos contendo oxigênio na borda ou superfície podem aumentar a transferência de elétrons. Os óxidos dopados de índio e estanho foram sintetizados pelo método dos precursores poliméricos (Pechini). O rGO foi obtido pela exfoliação química do grafite (método de Hummers modificado) seguida de redução com bissulfito de sódio. Os eletrocatalisadores contendo platina foram sintetizados pelo método de redução por borohidreto de sódio. Os pós cerâmicos particulados de ITO e ATO foram calcinados em diferentes temperaturas. A calcinação a 450 °C resultou no melhor suporte para o catalisador de platina, frente ao carbono Vulcan XC- 72, mais utilizado, segundo a literatura, no que se refere à reação de oxidação do etanol. A inclusão do rGO no suporte mostrou-se mais efetiva quando este é submetido a tratamento de agitação laminar de alta energia para a redução do tamanho das folhas. Porém estudos devem ser realizados para melhorar seu desempenho eletroquímico. Comparando-se todos os suportes estudados, observou-se que os compósitos Pt/ITO e Pt/ATO apresentaram os melhores resultados para a eletrooxidação do etanol. Observou-se que o aumento da área superficial dos pós e o efeito bifuncional promovido pelos óxidos são fatores importantes em sua aplicação como eletrodo.

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  • IPEN-DOC 26626

    CAMARGO, VICTOR F. de . Síntese de eletrocatalisadores de PtRh/C-ITO pelo método de borohidreto de sódio para eletrooxidação do etanol em meio alcalino / Synthesis of PtRh/C-ITO electrocatalysts prepared with sodium borohidride method for ethanol electrooxidation in alkaline media . 2019. Dissertação (Mestrado em Tecnologia Nuclear) - Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP, São Paulo. 80 p. Orientador: Almir Oliveira Neto. DOI: 10.11606/D.85.2020.tde-27122019-133317

    Abstract: Eletrocatalisadores de PtRh/C-ITO foram preparados em uma única etapa, usando H2PtCl6.6H2O e RhCl3.xH2O como fonte dos metais, borohidreto de sódio como agente redutor e uma mistura física de 85% de carbono Vulcan XC-72 e 15% In2O3.SnO2 (indium tin oxide - ITO) como suporte. PtRh/C-ITO preparados neste trabalho foram caracterizados por difração de raios X (DRX), microscopia eletrônica de transmissão (MET), espectroscopia de fotoelétrons excitados por raios X (XPS), espectroscopia in situ de infravermelho com transformada de Fourier (ATR-FTIR), voltametria cíclica, cronoamperometria e testes de performance em uma célula a combustível de etanol direto (DEFC). Espectros de difração de raios X para todos eletrocatalisadores de PtRh/C-ITO indicaram um deslocamento nos picos da Pt(fcc), mostrando que o Rh foi incorporado na matriz da Pt. Histogramas obtidos pelas imagens do MET para PtRh/C-ITO mostraram nanopartículas, com tamanho entre 3,0 e 4,0 nm, homogeneamente distribuídas sobre o suporte. Resultados do XPS da PtRh(70:30)/C-ITO mostraram a presença de uma mistura de espécies de diferentes estados de oxidação (Sn0 e SnO2), o que pode favorecer a oxidação de espécies intermediarias adsorvidas, através do mecanismo bifuncional. PtRh(90:10)/C-ITO foi a mais ativa nos estudos eletroquímicos devido a maior produção de CO2, indicando possuir maior seletividade na quebra da ligação C-C. Experimentos em DEFC mostraram que os valores de densidades de potência obtidas com PtRh(70:30)/C-ITO e PtRh(90:10)/C-ITO foram maiores do que com o Pt/C, indicando um efeito benéfico na adição de Rh a Pt, além do ITO no suporte de carbono.

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  • IPEN-DOC 26625

    YOSHIMURA, TANIA M. . Fotobiomodulação na síndrome metabólica : efeitos nos tecidos adiposos branco e marrom de camundongos / Photobiomodulation in metabolic syndrome: effects on white and brown adipose tissues from mice . 2019. Tese (Doutorado em Tecnologia Nuclear) - Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP, São Paulo. 90 p. Orientador: Martha Simões Ribeiro. DOI: 10.11606/T.85.2020.tde-12122019-172926

    Abstract: A síndrome metabólica (SM) é uma condição clínica que agrupa uma variedade de morbidades, como intolerância à glicose e obesidade. Na obesidade, o tecido adiposo branco (TAB) apresenta características inflamatórias que interferem na ação da insulina, levando à ocorrência de Diabetes do tipo 2. O tecido adiposo marrom (TAM), que tem como principal função a termogênese através da oxidação mitocondrial de cadeias carbônicas, se encontra hiporresponsivo aos estímulos clássicos na SM, como, por exemplo, a exposição ao frio. Estratégias para modular os processos inflamatórios do TAB e ativar o metabolismo do TAM podem atenuar as consequências da SM. Os reconhecidos efeitos anti-inflamatórios e de ativação do metabolismo mitocondrial da fotobiomodulação (PBM) indicam que essa poderia ser uma proposta terapêutica para a SM. Sendo esse o nosso foco de estudo, camundongos adultos, machos, da linhagem C57BL/6 receberam dieta hiperlipídica para indução da SM. Os animais foram então irradiados usando um dispositivo LED sobre a superfície abdominal (λ = 850 nm) ou interescapular (λ = 660 nm) para modular a inflamação do TAB ou ativar o TAM, respectivamente. O tratamento consistiu em 6 sessões de irradiação, distribuídas no decorrer de 21 dias. Apesar de não terem apresentado alterações na massa corporal e Índice de Lee, os animais irradiados na região abdominal (HFABD850) apresentaram 50 % menos células inflamatórias no TAB epididimal e também apresentaram melhora no teste de tolerância à glicose 24 h após a última sessão de tratamento. Nos animais obesos irradiados na região interescapular (HFTAM660), as irradiações promoveram aumento de duas vezes na massa do TAM, além de aumento da temperatura dorsal e da captação de 18F-FDG após exposição a baixas temperaturas. O soro desses animais (HFTAM660) também se mostrou mais semelhante ao de animais eutróficos. Nossos achados indicam que a PBM, nos parâmetros investigados, pode ser aplicada ao tratamento da SM.

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  • IPEN-DOC 26624

    GONÇALVES, PEDRO do N. . Caracterização química inorgânica e distribuição vertical de radionuclídeos das séries de decaimento do 238U e 232Th e 40K em testemunhos de sedimento e perfis de solo coletados na área de influência do reservatório de Jundiaí, estado de São Paulo / Inorganic chemical characterization and vertical distribution of natural radionuclides from 238U and 232Th series and 40K determined in sediment cores and soil profiles collected in the catchment area of Jundiaí reservoir, state of São Paulo . 2019. Dissertação (Mestrado em Tecnologia Nuclear) - Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP, São Paulo. 152 p. Orientador: Sandra Regina Damatto. DOI: 10.11606/D.85.2020.tde-03122019-111205

    Abstract: O reservatório de Jundiaí, localizado no estado de São Paulo, é um dos reservatórios de água que integram o SPAT - Sistema Produtor do Alto Tiête. A presença de poluentes no solo e no sedimento do manancial é um dos parâmetros para a avaliação da contaminação ambiental, que por sua vez pode afetar a qualidade da água do reservatório. O objetivo desta dissertação foi determinar as concentrações de elementos maiores e traços e as concentrações de atividade de radionuclídeos naturais em perfis de solo e testemunhos de sedimento do reservatório de Jundiaí. Os elementos traço As, Br, Co, Cr, Cs, Hf, Rb, Sb, Sc, Ta, Cd, La, Ce, Nd, Sm, Eu, Tb, Yb, Lu e Se e os elementos maiores Fe, K e Na foram determinados por Análise por Ativação com Nêutrons Instrumental (INAA); os radionuclídeos naturais das séries de decaimento do 238U e 232Th e o radionuclídeo 40K foram determinados por espectrometria gama. Avaliou-se também o fator de enriquecimento dos elementos maiores e traço utilizando os valores de referência de concentração na Crosta Continental Superior (CCS). Os parâmetros físico-químicos das amostras de solo e sedimento foram determinados com o intuito de verificar a influência que eles desempenham na disponibilidade dos radionuclídeos e elementos traço no solo e sedimento. Os elementos As e Br apresentaram enriquecimento que variaram de moderado até significante nos perfis de solo e testemunhos de sedimento. O elemento Se apresentou enriquecimento significante nos três testemunhos de sedimento analisados; as concentrações médias obtidas nos três testemunhos foram 5,4 mg.kg-1, 2,4 mg.kg-1 e 2,2 mg.kg-1. Esse elemento é considerado um elemento potencialmente tóxico (EPT) e pode ocasionar efeitos adversos à biota quando em concentrações elevadas. O radionuclídeo 232Th apresentou valores de concentração de atividade que ultrapassaram os valores de referência do UNSCEAR (United Nations Scientific Committee on the Effects of Atomic Radiation) em todos os compartimentos analisados; os radionuclídeos 238U e 226Ra também apresentaram valores mais altos que os níveis de referência do UNSCEAR em parte dos perfis e testemunhos analisados.

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  • IPEN-DOC 26623

    SOUZA, P.R.D. de . Avaliação comparativa de dosimetria com LiF:Mg,Ti (TLD-100) em phantom antropomórfico com o sistema de planejamento (TPS) para câncer de pulmão / Comparative assessment of LiF:Mg, Ti (TLD-100) dosimetry in anthropomorphic phantom and planning system (TPS) for lung cancer . 2019. Dissertação (Mestrado em Tecnologia Nuclear) - Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP, São Paulo. 70 p. Orientador: Maria Elisa Chuery Martins Rostelato. DOI: 10.11606/D.85.2020.tde-09122019-121029

    Abstract: O câncer de pulmão é o mais comum de todos os tumores malignos. Em 90% dos casos diagnosticados o câncer de pulmão esta associado ao consumo de derivados de tabaco. A radioterapia atua como forma de tratamento e existe duas formas de aplicação; a teleterapia e a braquiterapia. Na teleterapia é utilizado um acelerador linear para fazer a aplicação da dose. Antes de começar o tratamento é realizado um planejamento que faz a aquisição de todas informações anatômicas do paciente e em seguida a classificação das áreas de interesse para o tratamento. Na radioterapia a dosimetria é aplicada como uma forma de medição independente e neste trabalho tem como objetivo fazer a comparação do plano dosimétrico de tratamento com os valores de dose calculados no sistema de planejamento (TPS) utilizando um phantom antropomórfico. A dosimetria foi realizada com dosímetros termoluminescentes (Lif:Mg,Ti-TLD-100). Foram selecionados 25 TLD's que passaram por processo de seleção com as seguintes etapas: tratamento térmico, seguido de irradiação, leitura e posteriormente a calibração para uso no acelerador linear. Com os dosímetros já selecionados, foi elaborado o plano de tratamento feito no sistema de planejamento Eclipse da Varian e em seguida comparado à dosimetria realizada com os TLD'S alocados no phantom antropomórfico, para este mesmo caso. Um acelerador linear com energia de fótons/6MV, modelo 2100 da Varian foi utilizado para fazer a aplicação da dose de 200 cGy e 250 cGy. Os valores obtidos apresentaram-se de acordo com o recomentado pelos protocolos, 5% AAPM-TG-51 e 5 a 7% ICRU 50 e 60.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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).

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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  • IPEN-DOC 26378

    GOMES, DANIEL de S. ; GIOVEDI, CLAUDIA. Combining probabilistic and deterministic methods for accident analysis. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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  • IPEN-DOC 26377

    GOMES, DANIEL de S. ; GIOVEDI, CLAUDIA. Comparative analysis of silicon carbide with zirconium-based alloys. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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  • IPEN-DOC 26372

    LOBO, RAQUEL de M. ; ANDRADE, ARNALDO H.P. de . Advancesin the understanding of the mechanisms of iodine-induced SCC cracking in zirconium alloys. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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  • IPEN-DOC 26361

    GOMES, DANIEL de S. ; GIOVEDI, CLAUDIA. Importance of uncertainty modelling for nuclear safety analysis. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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>.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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 inteligent processing. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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  • IPEN-DOC 26352

    SILVA, LEANDRO G.M. e ; SABUNDJIAN, GAIANE . Virtual visit to nuclear research reactor IEA-R1. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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  • IPEN-DOC 26336

    OLIVEIRA, OTAVIO L. de ; BITELLI, ULYSSES D. . Future challenges for IPEN/MB-01 nuclear research reactor. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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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: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 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.

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  • IPEN-DOC 26332

    SAVOINE, MARCIA M. ; ANDRADE, DELVONEI A. ; MENEZES, MARIO O. de . Methodology proposal for assessing safety in WSN and IoT devices in nuclear research laboratory. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3827-3837.

    Abstract: Nowadays there is a gap due to the absence of an updated and formalized methodology that can be used to assess security levels in WSNs (Wireless Sensor Network) under IoT (Internet of Things) devices in nuclear environments (which are considered hostile environments and require a higher level of security). This gap causes information security professionals to have di culties in making a broad assessment of the vulnerabilities in their WSNs, with greater concern when coupled with IoT devices. This work aims to present a methodology to evaluate the reliability of the use of levels security with IoT devices for nuclear installations using WSNs. The proposal of the methodology consists of 5 main stages and 21 substages, which are part of the category of a function in groups of cyber security results that are linked to programmatic needs and speci c activities of mandatory execution. Understanding so that the security of a WSN considering the current IoT context for nuclear installations is necessary, where important characteristics in these critical environments should be explored (e g., the presence of radioactivity, in addition to the decontamination of materials and equipment, determine access to authorized persons). The application of the defense-in-depth concept of anomaly solution management and prevention against atypical events to provide an e ective safety mechanism, ensuring its safe use in these high criticality environments.

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  • IPEN-DOC 26331

    BARABAS, ROBERTA de C. ; BARABÁS, CARLOS ; SABUNDJIAN, GAIANE . The development of a multisensory program for the dissemination of the beneficial applications of the nuclear technology. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3770-3781.

    Abstract: Despite all peaceful applications of nuclear technology, it is still addressed with prejudice. Prejudices may be explicit (conscious) or implicit (unconscious). However, either explicit or implicit, they interfere with individuals’ behavior and attitudes. Prejudices against any theme may be reduced and even reversed by new learning on the theme. Multisensory techniques have proven to make learning richer and more motivating. This work aims to present the development of a multisensory program designed for learning about the beneficial applications of nuclear technology and compare this program to a 12-week traditional teaching program with lecture classes about the nuclear technology. The multisensory program was held at the Instituto de Pesquisas Energéticas e Nucleares (IPEN) for a group of teachers. Assisted tours to the IEA-R1 and to the Centro da Tecnologia das Radiações (CTR) as well as a coffee break serving a variety of commercially-available foods containing irradiated ingredients were part of the multisensory approach. The Implicit Association Test (IAT) was administered before and after the program to identify and measure the implicit associations towards the nuclear technology. This multisensory program was compared to a 12-week traditional teaching program with lecture classes about the nuclear technology held at IPEN. Unlike the multisensory program, the IAT results from the traditional program demonstrated that the lecture classes were not effective for changing the implicit associations. The multisensory program was an effective tool for changing the implicit associations and can be useful for disseminating the beneficial applications of the nuclear technology.

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  • IPEN-DOC 26330

    FREIRE, LUCIANO O. ; ANDRADE, DELVONEI A. de . Entering new markets: nuclear industry challenges. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3714-3722.

    Abstract: Nuclear ship propulsion and isolated islands energy supply are unexplored markets for nuclear vendors. Carbon taxes and fuel regulations may make fossil fuels more expensive. Such markets pay more for energy because of organization and transport costs and use of small machines, which are less efficient than grid generators. The goal of this work is to find the measures the nuclear industry needs to take to get into new potential markets. This work shows the different actors and their interests and points the natural or physical constraints they face. Considering interests and constraints, this work named the most probable market niches where nuclear power may beat other power sources. After considering natural constraints, this paper analyses human-generated constraints and presents a way on how to mitigate or solve them. This study shows that nuclear industry needs to take technical, administrative, and political measures before nuclear power arrives to a wider market. This work is based on literature review and qualitative analysis and cannot point precise thresholds where nuclear power should be competitive. Future work will consist of statistical analysis to find precise thresholds to help in the decision-making process.

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  • IPEN-DOC 26329

    SMITH, RICARDO B. ; SACHDEVA, MAHIMA; BISURI, INDRANIL; VICENTE, ROBERTO . Advanced heavy water reactor: a new step towards sustainability. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3567-3579.

    Abstract: One of the great advances in the current evolution of nuclear power reactors is occurring in India, with the Advanced Heavy Water Reactor (AHWR). It is a reactor that uses thorium as part of its fuel, which in its two fueling cycle options, in conjunction with plutonium or low enriched uranium, produces energy at the commercial level, generating less actinides of long half-life and inert thorium oxide, which leads to an optimization in the proportion of energy produced versus the production of burnt fuels of the order of up to 50%. The objective of this work is to present the most recent research and projects in progress in India, and how the expected results should be in compliance with the current sustainability models and programs, especially the "Green Chemistry", a program developed since the 1990s in the United States and England, which defines sustainable choices in its twelve principles and that can also be mostly related to the nuclear field. Nevertheless, in Brazil, for more than 40 years there has been the discontinuation of research for a thorium-fueled reactor, and so far there has been no prospect of future projects. The AHWR is an important example as an alternative way of producing energy in Brazil, as the country has the second largest reserve of thorium on the planet.

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  • IPEN-DOC 26328

    COELHO, ADRIANA B.; CONTI, THADEU N. . Measurement of the generation of electrical energy in a photovoltaic system grid-connected in the Amazon region in the rain period. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3517-3531.

    Abstract: Since 2012, when Resolution No. 482 of ANEEL (National Agency for Electric Energy) created the Electric Energy Compensation System, it was possible for Brazilian consumers to generate their own electricity from renewable sources or qualified cogeneration, supply the surplus to the distribution network of your locality. This milestone motivated the industry to develop technology in the area of photovoltaic energy. In light of this new perspective, the objective of this article is to compare the generation of electric energy by Grid-Connected Photovoltaic Power System 3.1 kWp installed in the rural area of the State of Rondônia located in the Amazon region, where the climatic seasons are rain and dry, with the generation estimate of the PVSyst program. The results of this analysis suggest that the industry develop projects and research to improve the program when it involves grid-connected photovoltaic (PV) power system in the northern region.

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  • IPEN-DOC 26327

    MOREIRA, RENAN P. ; TATEI, TATIANE Y. ; ARAUJO, DANIELLE G.; DUQUE, MARCO A. da S.; OLIVEIRA, IVAN C. de; AYOUB, JAMIL M.S. ; SENEDA, JOSE A. . Prospects for nuclear energy in Brazil. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3511-3516.

    Abstract: One of the main purposes of nuclear technology is to produce electricity, with the advantage of producing a lower volume of radioactive waste. The expansion of nuclear energy in the electrical system has been positive, as it is one of the types of energy that is available at any time and in the desired amount. Considered a reliable source and safe alternative to compose a country's energy matrix. In the case of Brazil, it has enough reserves of Uranium and Thorium to compose the energy matrix over many years. The increase in demand, and the need for energy from renewable sources has caused changes in the world's electric power generation. According to World Nuclear Association (WNA), 14% of the energy is generated by nuclear energy sources, and this percentage tends to increase with the construction of new plants. According to the International Atomic Energy Agency (IAEA), the goal for nuclear energy is to provide 25% of electricity in 2050. Other technologies are applied in the nuclear area, for example nuclear medicine, in which radioactive materials are used with low doses of radiation for treatment and diagnosis of diseases, even in development are effective and safe, especially in the areas of cardiological, neurological and oncological diagnosis. Despite the knowledge acquired with the development of Brazilian nuclear projects, many are partly lost and discontinuity investments of successive governments, therefore, this work intends to study an overview of nuclear energy in Brazil in recent years and its prospects.

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  • IPEN-DOC 26326

    FRENZEL, LUCAS S. ; SABUNDJIAN, GAIANE . Análise teórico/experimental do fenômeno de circulação natural no circuito de circulação natural do IPEN. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3433-3444.

    Abstract: O objetivo deste trabalho é o estudo do fenômeno de circulação natural em circuitos experimentais para aplicação em instalações nucleares. Trabalhos sobre circuitos de circulação natural ganharam força após o acidente de Three Mile Island. Este acidente mostrou que a segurança deste tipo de reator não era suficientemente confiável. Outro ponto importante é relacionado a necessidade de intervenção humana para a entrada de operação dos sistemas de segurança, evidenciando que erros operacionais foram as maiores causas para o acidente de Three Mile Island. Assim, há um crescente interesse da comunidade científica no estudo da circulação natural devido ao seu uso na nova geração de reatores nucleares compactos. O circuito experimental utilizado neste estudo foi reparado/ modernizado, e se encontra no Centro de Engenharia Nuclear do Instituto de Pesquisas Energéticas e Nucleares (CEN-IPEN). Para a realização deste trabalho, foi simulado alguns experimentos com diferentes: níveis de potência e vazão de água no secundário; originando um banco de dados experimentais que é utilizado para validar alguns programas termohidráulicos. Particularmente para este estudo, os resultados experimetais obtidos são comparados com o modelo teórico criado com o código RELAP/MOD3.3 [1]. Os resultados obtidos com o programa são satisfatórios quando comparados com os experimentais.

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  • IPEN-DOC 26325

    SILVESTRE, LARISSA J.B. ; SOUSA, EMERSON L.; SABUNDJIAN, GAIANE . Pós-processador matemático para o software de teste de associação implicita – FreeIAT. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3368-3374.

    Abstract: Uma das formas de identificar algum tipo de preconceito é por meio do uso de softwares de técnicas de neurociências aplicadas ao uso de medida da memória implícita (Testes de Associação Implícita –TAI), que não depende da atenção consciente do participante, sendo suas respostas automáticas e espontâneas. Os seguintes testes de associação implícita foram encontrados na literatura: o Teste de Associação Implícita, o Priming, o Visual Organization Test (VOT) e o Inquisit. Dentre todos os softwares de associação implícitas apresentados, o FreeIAT será utilizado neste trabalho pelo fato de ser um programa largamente usado e validado em diversas pesquisas. Pelo fato desse programa apresentar resultados bem consistentes quanto à identificação de possíveis preconceitos em vários temas, viu-se a necessidade de elaborar um pós-processador matemático a fim de automatizar os resultados em forma de gráficos. Portanto, o objetivo desse trabalho é o de desenvolver um pós-processador matemático com interface amigável, que facilitará a apresentação e interpretação dos resultados dos usuários do FreeIAT e poderá ser utilizado em qualquer área de interesse. A linguagem utilizada para o desenvolvimento desse pós-processador é o C#. Os resultados preliminares desse novo pós-processador mostraram-se eficientes.

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  • IPEN-DOC 26324

    DIAS, ANDRESSA de J.R. ; VICENTE, ROBERTO ; DELLAMANO, JOSE C. . Analysis of accidents in industrial gammagraphy. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3199-3205.

    Abstract: This study presents industrial gammagraphy accidents from 1967 to 2015, as a way to help the improvement of knowledge to radiation protection and the prevention of futures accidents, based on its common causes. It is based on a research in progress. The term radiation protection is applied to the concept of protection of people, worker or public, against the harmful effect of ionizing radiation. It is an important area and has to be in constant improvement to gain the society’s trust. A way to make it possible is through studies of past accidents therefore, accidents reports are important. It is useful for creating a database with enough information to assist in accident management and prevention. This database also helps radiation practices to be more accepted by the community. From a public individual point of view, a practice with reliable statistics that shows low accident rates is more acceptable, even though some hazard might be present. The intent is gammagraphy’s risks to be managed and reduced in the future, so the use of the technology might grow while public’s acceptance increases and the magnitude of the perceived danger of the practice diminishes as seen through people’s eyes.

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  • IPEN-DOC 26323

    OLIVEIRA, VITORIA A. ; CARVALHO, ELITA U. ; DURAZZO, MICHELANGELO ; SAKATA, SOLANGE K. ; GARCIA, RAFAEL H.L. . Adsorção líquida no siliceto de urânio. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3193-3198.

    Abstract: O siliceto de urânio é um intermetálico usado como combustível nuclear na maioria dos reatores de pesquisa modernos, incluindo os reatores MB-01 e IEA-R1 do IPEN. Durante a produção, o material é submetido a um rigoroso controle de qualidade, que inclui análises de tamanho de partícula, densidade, caracterização e composição da fase cristalina. A quantificação das fases cristalinas presentes é realizada por difração de raios X (DRX) e refinamento dos dados usando o método Rietveld. No entanto, devido à alta absorção de raios X por esse material, no que diz respeito ao método de quantificação adotado, é muito importante reduzir o tamanho das partículas. Para este objetivo, um moinho vibratório dedicado é usado antes da análise de DRX, reduzindo o diâmetro médio das partículas para poucos micrômetros. Para evitar a oxidação das amostras, o processo de moagem ocorre em meio isopropanóico, o qual é seco posteriormente, em vácuo a 80 ºC. Porém, em muitos casos, verifica-se que as massas das amostras moídas são maiores do que as iniciais. Nesse sentido, esse trabalho propõe analisar a causa dessa diferença de massa. Granulometria a laser, termogravimetria (TG). Os resultados de TG sugerem que uma camada é fortemente adsorvida ao material, protegendo o pó de oxidação em temperaturas acima de 4000C.

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  • IPEN-DOC 26322

    MIURA, VINICIUS T.; ZAMBONI, CIBELE B. ; GIOVANNI, DALTON N. S. ; SANTOS, PAULA A.D.A. de S. ; RIZZUTTO, MARCIA A.. Sua foto é um documento histórico?. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3180-3185.

    Abstract: Neste estudo a técnica de Fluorescência de Raios X por Dispersão de Energia (FRXDE) foi utilizada para a investigação de uma coleção fotográfica “Palacetes de São Paulo”, constituída por 48 fotos, cuja data e processo de produção não são conhecidos. A coleção faz parte de um acervo particular e foi disponibilizada para as análises no Laboratório de Espectroscopia e Espectrometria das Radiações (IPEN/CNEN-SP). A presença majoritária de Ba, bem como a presença de S, Cl, Ca, Fe, Sr e Au (em menor teor) identificados pela técnica em todas as fotos, é coerente com processo de revelação que utiliza papel fotográfico com revestimento de Barita (BaSO4), viragem de Au (para preservação) e fixadores a base de cloretos (CaCl2 e FeCl3). Este papel fotográfico foi introduzido no mercado em 1894 e muito utilizado por fotógrafos profissionais e amadores até meados de 1930, quando deixou de ser comercialmente produzido. Esses resultados fornecem aos colecionadores / conservadores subsídios para o correto armazenamento e preservação. Ainda, para Historiadores e Curadores agregam conhecimento de relevância histórica aos acervos fotográficos e compõem informações fundamentais para museus (catalogação / registro), particularmente no que diz respeito `a arquitetura paulistana, ampliando seu conhecimento bem como para a realização de exposições. Para fotógrafos profissionais agregam conhecimento no âmbito técnico.

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  • IPEN-DOC 26321

    SOUZA, ERIC W. de ; VIEIRA, JOSE M. ; SILVA, LEONARDO G. de A. e . Uso da radiação ionizante na reciclagem de poli (tetrafluoroetileno) (PTFE). In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3167-3171.

    Abstract: A maioria dos países enfrenta grandes desafios para controlar e organizar a geração e a disposição dos resíduos sólidos urbanos. Milhões de toneladas desses resíduos são gerados anualmente pela população e pelas indústrias. A eliminação destes resíduos sólidos é um problema mundial crescente. Os materiais poliméricos (plásticos e borrachas) compreendem uma proporção cada vez maior de resíduos industriais que entram em aterros sanitários e ambientais. Devido à capacidade da radiação ionizante alterar a estrutura e as propriedades dos materiais poliméricos, e o fato de que ela é aplicável a todos os tipos de polímeros, a irradiação é promissora para tratar do problema de resíduos poliméricos. O objetivo deste trabalho foi utilizar a radiação gama proveniente de uma fonte de 60Co para reciclar o poli(tetrafluoroetileno) (PTFE) que é um polímero de difícil decomposição quando descartado no meio ambiente. Aparas industriais deste polímero foram selecionadas e submetidas ao processo de moagem. Posteriormente, as amostras foram submetidas ao processo de irradiação com uma dose de 200 kGy. Após a irradiação o material obtido foi micronizado obtendo-se um pó muito fino de PTFE o qual foi classificado de acordo com os tamanhos de partículas com características especiais para diferentes possibilidades de utilização industrial (aditivos para tintas, massas lubrificantes, óleos e como carga em polímero para diminuir o coeficiente de atrito).

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  • IPEN-DOC 26320

    SILVA, CAMILA L. ; TOMINAGA, FLAVIO K. ; JACOVONE, RAYNARA M.S. ; BRANDAO, OCTAVIO A.B. ; SAKATA, SOLANGE K. . Estudo de estabilidade de nanocompósitos de magnetita/óxido de grafeno reduzido sintetizados via feixe de elétrons. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3158-3166.

    Abstract: O óxido de grafeno é um dos precursores do grafeno e apresenta em sua superfície vários grupos funcionais oxigenados que consequentemente possui dispersibilidade em diversos solventes polares, o que lhe proporciona alta competência para em diversas aplicações. Este nanomaterial possui excelentes propriedades físico-químicas, como estabilidade mecânica, mobilidade elétrica, condutividade térmica. A solubilidade pode ser aprimorada por meio da formação de uma barreira estérea quando disperso em água, que causa então a diminuição das interações eletroestática entre as partículas. Diversos metais têm sido incorporados a nanocompósitos a base de grafeno. A síntese de nanocompósitos de óxido grafeno/magnetita tem sido estudada devido ao aumento das propriedades magnéticas, catalíticas e da biocompatibilidade. Este trabalho tem como finalidade avaliar a estabilidade de nanocompósitos magnéticos de óxido de grafeno obtidos através da irradiação com feixes de elétrons. Os nanocompósitos foram irradiados em um acelerador de elétrons em diferentes doses (20, 40 e 80 kGy). Os métodos de caracterização usados foram espectrofotometria UV/VIS e potencial zeta (ζ). Nas análises de UV/VIS foi observado o pico padrão de absorção na região de 230nm, o que confirma a existência de ligações C=C. As análises do potencial zeta foram realizadas nos pH de 4, 7 e 9 e a maior estabilidade foi obtida em pH 7 nas amostras irradiada a 20 kGy e 80 kGy.

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  • IPEN-DOC 26319

    ANDRADE, MARIANA N. ; OLIVEIRA, GLAUCIA A.C. ; PIRANI, DEBORA A. ; COUTINHO, JOAO F. ; BERGAMASCHI, VANDERLEI S. ; SENEDA, JOSE A. ; BUSTILLOS, JOSE O.V. . Purification of lithium carbonate by ion-exchange processes for application in nuclear reactors. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3153-3157.

    Abstract: Lithium Compounds have applications in strategic areas for intern consumption of a country as well as international commerce. In nuclear industry, the lithium is used for the cooling of PWR reactors as a pH stabilizer. Based on this assumption, the generation of knowledge to master the processing cycle of these compounds is essential. The high degree of purity of lithium compounds is determinant to have success in these applications. Lithium hydroxide LiOH and lithium carbonate Li2CO3 are the main forms in which lithium is used industrially. To improve the quality of the starting product, purifying process were used until obtaining an adequate purity level of raw material (> 99%). The present work aims to make feasible a purification of Li2CO3 through ion-exchange chromatography from a 98.5% purity compound. The impurities present in higher content are sodium and calcium. To separate these two elements from lithium or at least to lower their concentrations, a column with cationic resin was used to fix lithium. The determination of lithium, sodium and calcium contents in the solutions was performed by inductively coupled plasma optical emission spectrometry, ICP-OES. The experiments performed to evaluate the best lithium purification condition were based on the variation of the main operational parameters: pH, flow and elution solution. The results indicate increased purity from the application of ion exchange operations obtaining a suitable condition for nuclear uses.

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  • IPEN-DOC 26318

    BRANDAO, OCTAVIO A.B. ; SILVA, CAMILA L. da ; JACOVONE, RAYNARA M.S. ; TOMINAGA, FLAVIO K. ; SAKATA, SOLANGE K. . Estudo de estabilizantes para o óxido de grafeno reduzido por radiação gama. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3144-3152.

    Abstract: O estudo do óxido de grafeno (OG) e de nanocompósitos à base do óxido de grafeno mostra se relevante devido à sua versatilidade em inúmeras aplicações, como na síntese de biossensores e na adsorção de nanopartículas metálicas. Por ser um nanomaterial, há uma grande dificuldade de impedir sua aglomeração em meio aquoso, gerando a necessidade de uma melhor compreensão de sua estabilidade Este trabalho propõe se a realizar um estudo da estabilidade do óxido de grafeno reduzido por radiação gama em diferentes dispersões: em meio aquoso básico, em poliacetato de vinilia (PVA), propano 2 ol, etileno glicol (EG) e água. Os métodos empregados para a caracterização do óxido de grafeno foram o DRX, FTIR e UV Vis. Os resultados obtidos na análise dos espectros das amostras irradiadas pelo DRX indicaram que houve a redução nas dispersões com Água, ISO, EG e PVA pelo deslocamento do pico característicos do OG de 10º nestas amostras, corroborado pelo deslocamento do pico da absorbância do UV Vis para a faixa de 240 270 nm, entretanto a amostra de NaCl não reduziu conforme visto no FTIR. Em meio aquoso houve uma redução na intensidade dos picos indicando a aglomeração do nanomaterial com o decorrer do tempo de análise. O uso dos estabilizantes ISO, PVA e EG melhor minimizaram este processo

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  • IPEN-DOC 26317

    PEREIRA, DEBORA A.; FERREIRA, DOUGLAS A.; FATTE, MARIO; SOUZA, NATALIA DE O.; GIOVEDI, CLAUDIA; COTRIM, MARYCEL E.B. ; PIRES, MARIA A. . Análise química de liga de grau nuclear aplicada como material de controle em reatores nucleares. In: ABEN (Ed.) INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro, 2019. p. 3117-3129.

    Abstract: A liga de prata-indio-cádmio (Ag/In/Cd) é utilizada como material absorvedor em elementos de controle de reatores nucleares devido à alta seção de choque para absorção de nêutrons de seus componentes. Em Reatores Refrigerados a Água Pressurizada (PWR - Pressurized Water Reactor), a liga Ag/In/Cd é utilizada na forma de barra contendo 80% de prata, 15% de índio e 5% de cádmio em massa com tolerâncias, máxima e mínima, bastante rigorosas em sua composição. A liga na forma de barra é encapsulada em tubos metálicos, os quais compõem o conjunto do elemento de controle no reator nuclear. Para ser aplicada com este propósito, a barra de liga Ag/In/Cd deve apresentar uma composição homogênea ao longo de toda a sua extensão, a fim de assegurar seu comportamento adequado dentro do reator. O objetivo deste projeto é desenvolver e qualificar a metodologia de análise química aplicada à caracterização da liga Ag/In/Cd para ser usada em barras de controle em reatores do tipo PWR. A metodologia padronizada para determinar o teor de prata, índio e cádmio na liga de grau nuclear é a titulação potenciométrica para prata e a titulação de complexação para o índio e o cádmio. A precisão dos resultados obtidos depende da prévia calibração dos materiais volumétricos e equipamentos utilizados, bem como da calibração dos reagentes titulantes a serem utilizados na titulação. Além disso, a qualificação desse processo para fins nucleares requer a elaboração de todos os documentos relacionados a cada uma das etapas do processo, incluindo práticas operacionais e registros da qualidade. O desenvolvimento e a qualificação da metodologia representam passos fundamentais no sentido de tornar o Brasil autossuficiente na produção desse material aplicado à área nuclear.

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A pesquisa no RD utiliza os recursos de busca da maioria das bases de dados. No entanto algumas dicas podem auxiliar para obter um resultado mais pertinente.

É possível efetuar a busca de um autor ou um termo em todo o RD, por meio do Buscar no Repositório , isto é, o termo solicitado será localizado em qualquer campo do RD. No entanto esse tipo de pesquisa não é recomendada a não ser que se deseje um resultado amplo e generalizado.

A pesquisa apresentará melhor resultado selecionando um dos filtros disponíveis em Navegar

Os filtros disponíveis em Navegar tais como: Coleções, Ano de publicação, Títulos, Assuntos, Autores, Revista, Tipo de publicação são autoexplicativos. O filtro, Autores IPEN apresenta uma relação com os autores vinculados ao IPEN; o ID Autor IPEN diz respeito ao número único de identificação de cada autor constante no RD e sob o qual estão agrupados todos os seus trabalhos independente das variáveis do seu nome; Tipo de acesso diz respeito à acessibilidade do documento, isto é , sujeito as leis de direitos autorais, ID RT apresenta a relação dos relatórios técnicos, restritos para consulta das comunidades indicadas.

A opção Busca avançada utiliza os conectores da lógica boleana, é o melhor recurso para combinar chaves de busca e obter documentos relevantes à sua pesquisa, utilize os filtros apresentados na caixa de seleção para refinar o resultado de busca. Pode-se adicionar vários filtros a uma mesma busca.

Exemplo:

Buscar os artigos apresentados em um evento internacional de 2015, sobre loss of coolant, do autor Maprelian.

Autor: Maprelian

Título: loss of coolant

Tipo de publicação: Texto completo de evento

Ano de publicação: 2015

Para indexação dos documentos é utilizado o Thesaurus do INIS, especializado na área nuclear e utilizado em todos os países membros da International Atomic Energy Agency – IAEA , por esse motivo, utilize os termos de busca de assunto em inglês; isto não exclui a busca livre por palavras, apenas o resultado pode não ser tão relevante ou pertinente.

95% do RD apresenta o texto completo do documento com livre acesso, para aqueles que apresentam o significa que e o documento está sujeito as leis de direitos autorais, solicita-se nesses casos contatar a Biblioteca do IPEN, bibl@ipen.br .

Ao efetuar a busca por um autor o RD apresentará uma relação de todos os trabalhos depositados no RD. No lado direito da tela são apresentados os coautores com o número de trabalhos produzidos em conjunto bem como os assuntos abordados e os respectivos anos de publicação agrupados.

O RD disponibiliza um quadro estatístico de produtividade, onde é possível visualizar o número dos trabalhos agrupados por tipo de coleção, a medida que estão sendo depositados no RD.

Na página inicial nas referências são sinalizados todos os autores IPEN, ao clicar nesse símbolo será aberta uma nova página correspondente à aquele autor – trata-se da página do pesquisador.

Na página do pesquisador, é possível verificar, as variações do nome, a relação de todos os trabalhos com texto completo bem como um quadro resumo numérico; há links para o Currículo Lattes e o Google Acadêmico ( quando esse for informado).

ATENÇÃO!

ESTE TEXTO "AJUDA" ESTÁ SUJEITO A ATUALIZAÇÕES CONSTANTES, A MEDIDA QUE NOVAS FUNCIONALIDADES E RECURSOS DE BUSCA FOREM SENDO DESENVOLVIDOS PELAS EQUIPES DA BIBLIOTECA E DA INFORMÁTICA.

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.

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.

1. Portaria IPEN-CNEN/SP nº 387, que estabeleceu os princípios que nortearam a criação do RDI, clique aqui.


2. A experiência do Instituto de Pesquisas Energéticas e Nucleares (IPEN-CNEN/SP) na criação de um Repositório Digital Institucional – RDI, clique aqui.