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Sample records for bolsa chica-2 reactor

  1. Bolsa Bay, California, Proposed Ocean Entrance System Study. Report 2. Comprehensive Shoreline Response Computer Simulation, Bolsa Bay, California

    Science.gov (United States)

    1990-04-01

    Southern California Bight is affected by a land-sea breeze pattern. A variation in flow is caused by the heating of the land surface during the day, and...1980). 27. The success of the inlet channel at Agua Hedionda indicates that a stable non-navigable entrance at Bolsa Chica could be feasible provided a...dual jetty system similar to Agua Hedionda is incorporated into the design. However, structures that penetrate into the active surf zone are expected

  2. When Informationists Get Involved: the CHICA-GIS Project.

    Science.gov (United States)

    Whipple, Elizabeth C; Odell, Jere D; Ralston, Rick K; Liu, Gilbert C

    2013-01-01

    Child Health Improvement through Computer Automation (CHICA) is a computer decision support system (CDSS) that interfaces with existing electronic medical record systems (EMRS) and delivers "just-in-time" patient-relevant guidelines to physicians during the clinical encounter and accurately captures structured data from all who interact with the system. "Delivering Geospatial Intelligence to Health Care Professionals (CHICA-GIS)" (1R01LM010923-01) expands the medical application of Geographic Information Systems (GIS) by integrating a geographic information system with CHICA. To provide knowledge management support for CHICA-GIS, three informationists at the Indiana University School of Medicine were awarded a supplement from the National Library Medicine. The informationists will enhance CHICA-GIS by: improving the accuracy and accessibility of information, managing and mapping the knowledge which undergirds the CHICA-GIS decision support tool, supporting community engagement and consumer health information outreach, and facilitating the dissemination of new CHICA-GIS research results and services.

  3. The Brazilian Bolsa Escola

    Directory of Open Access Journals (Sweden)

    Rachel Cassidy

    2008-12-01

    Full Text Available The Bolsa Escola (‘school stipend’ and its successor the Bolsa Familia (‘family stipend’ schemes have formed a crucial and successful part of Brazil’s welfare program. Bolsa Escola provided aid to Brazil’s poorest families on the condition that their children attended school, and Bolsa Familia has extended this idea, giving aid on the condition that children both attend school and receive vaccinations. Bolsa Familia is currently the largest Conditional Cash Transfer Program (CCTP in the world, costing roughly 0.5% of Brazilian GDP and helping around 11.2 million families (around 44 million Brazilians, constituting roughly one fifth of the population. Multilateral institutions have praised the schemes, and they are setting a leading example to other developing nations. In 2005, Paul Wolfowitz (former president of the World Bank said, ‘Bolsa Familia has already become a highly praised model of effective social policy. Countries around the world are drawing lessons from Brazil’s experience and are trying to produce the same results for their own people’.

  4. Antioxidant capacity of the leaf extract obtained from Arrabidaea chica cultivated in Southern Brazil.

    Directory of Open Access Journals (Sweden)

    Jackeline Tiemy Guinoza Siraichi

    Full Text Available Arrabidaea chica leaf extract has been used by people as an anti-inflammatory and astringent agent as well as a remedy for intestinal colic, diarrhea, leucorrhea, anemia, and leukemia. A. chica is known to be a good producer of phenolics. Therefore, in the present study, we investigated its antioxidant activity. The phenolic composition of A. chica leaves was studied by liquid chromatography coupled to diode array detection (LC-DAD and liquid chromatography coupled to electrospray ionization-tandem mass spectrometry (LC-ESI-MS/MS, and isoscutellarein, 6-hydroxyluteolin, hispidulin, scutellarein, luteolin, and apigenin were identified. The extract from leaves of A. chica was tested for antioxidant activity using the 2,2-diphenyl-1-picrylhydrazyl (DPPH method, β-carotene bleaching test, and total reactive antioxidant potential (TRAP method. The crude extract quenched DPPH free radicals in a dose-dependent manner, and the IC50 of the extract was 13.51 µg/mL. The β-carotene bleaching test showed that the addition of the A. chica extract in different concentrations (200 and 500 µg/mL prevented the bleaching of β-carotene at different degrees (51.2% ±3.38% and 94% ±4.61%, respectively. The TRAP test showed dose-dependent correlation between the increasing concentrations of A. chica extract (0.1, 0.5, and 1.0 µg/mL and the TRAP values obtained by trolox (hydro-soluble vitamin E 0.4738±0.0466, 1.981±0.1603, and 6.877±1.445 µM, respectively. The 2 main flavonoids, scutellarein and apigenin, were separated, and their antioxidant activity was found to be the same as that of the plant extract. These 2 flavonoids were quantified in the plant extract by using a validated HPLC-UV method. The results of these tests showed that the extract of A. chica had a significant antioxidant activity, which could be attributed to the presence of the mixture of flavonoids in the plant extract, with the main contribution of scutellarein and apigenin.

  5. Arrabidaea chica Hexanic Extract Induces Mitochondrion Damage and Peptidase Inhibition on Leishmania spp.

    Directory of Open Access Journals (Sweden)

    Igor A. Rodrigues

    2014-01-01

    Full Text Available Currently available leishmaniasis treatments are limited due to severe side effects. Arrabidaea chica is a medicinal plant used in Brazil against several diseases. In this study, we investigated the effects of 5 fractions obtained from the crude hexanic extract of A. chica against Leishmania amazonensis and L. infantum, as well as on the interaction of these parasites with host cells. Promastigotes were treated with several concentrations of the fractions obtained from A. chica for determination of their minimum inhibitory concentration (MIC. In addition, the effect of the most active fraction (B2 on parasite’s ultrastructure was analyzed by transmission electron microscopy. To evaluate the inhibitory activity of B2 fraction on Leishmania peptidases, parasites lysates were treated with the inhibitory and subinhibitory concentrations of the B2 fraction. The minimum inhibitory concentration of B2 fraction was 37.2 and 18.6 μg/mL for L. amazonensis and L. infantum, respectively. Important ultrastructural alterations as mitochondrial swelling with loss of matrix content and the presence of vesicles inside this organelle were observed in treated parasites. Moreover, B2 fraction was able to completely inhibit the peptidase activity of promastigotes at pH 5.5. The results presented here further support the use of A. chica as an interesting source of antileishmanial agents.

  6. Effect of Arrabidaea chica extracts on the Ehrlich solid tumor development

    Directory of Open Access Journals (Sweden)

    Ana Flávia C. Ribeiro

    2012-04-01

    Full Text Available The aim of this study was to investigate the effect of Arrabidaea chica (Humb. & Bonpl. B. Verl., Bignoniaceae, extracts on Ehrlich solid tumor development in Swiss mice. Leaves of A. chica were extracted with two distinct solvents, ethanol and water. The phytochemical analysis of the extracts indicated different classes of secondary metabolites like as anthocyanidins, flavonoids, tannins and saponins. Ethanol (EE and aqueous (AE extracts at 30 mg/kg reduced the development of Ehrlich solid tumor after ten days of oral treatment. The EE group presented increase in neutrophil count, α1 and β globulin values, and decrease of α2 globulin values. Furthermore, EE reduced the percentage of CD4+ T cells in blood but did not alter the percentage of inflammatory mononuclear cells associated with tumor suggesting a direct action of EE on tumor cells. Reduced tumor development observed in AE group was accompanied by a lower percentage of CD4+ T lymphocytes in blood. At the tumor microenvironment, this treatment decreased the percentage of CD3+ T cells, especially due to a reduction of CD8+ T subpopulation and NK cells. The antitumor activity presented by the AE is possibly related to an anti-inflammatory activity. None of the extracts produced toxic effects in animals. In conclusion, the ethanol and aqueous extracts of A. chica have immunomodulatory and antitumor activities attributed to the presence of flavonoids, such as kaempferol. These effects appear to be related to different mechanisms of action for each extract. This study demonstrates the potential of A. chica as an antitumor agent confirming its use in traditional popular medicine.

  7. bolsa

    Directory of Open Access Journals (Sweden)

    David Quintana Montero

    2007-01-01

    Full Text Available Este artículo aborda el fenómeno del rendimiento inicial de las salidas a bolsa a través de modelos que consideran la cuestión tanto desde un punto de vista longitudinal como transversal. La propuesta consiste en una forma de incorporar tanto la inercia del mercado primario como información relacionada con la estructura de la colocación al estudio de casos concretos. Los resultados ponen de manifiesto una mejora substancial de la capacidad explicativa de las regresiones empleadas.

  8. Chitosan–tripolyphosphate nanoparticles as Arrabidaea chica standardized extract carrier: synthesis, characterization, biocompatibility, and antiulcerogenic activity

    Directory of Open Access Journals (Sweden)

    Servat-Medina L

    2015-06-01

    Full Text Available Leila Servat-Medina,1,2 Alvaro González-Gómez,2,3 Felisa Reyes-Ortega,2 Ilza Maria Oliveira Sousa,1 Nubia de Cássia Almeida Queiroz,1 Patricia Maria Wiziack Zago,1 Michelle Pedrosa Jorge,1 Karin Maia Monteiro,1,4 João Ernesto de Carvalho,1 Julio San Román,2,3 Mary Ann Foglio1 1Chemical, Biological and Agricultural Pluridisciplinary Research Center-State University of Campinas (CPQBA-UNICAMP, Campinas-SP, Brazil; 2Biomaterials Group, Polymer Science and Technology Institute-Spanish National Research Council (ICTP-CSIC, 3CIBER-BBN, Centro de Investigación Biomédica en Red, Madrid, Spain; 4Department of Medical Clinics, Faculty of Medical Sciences, University of Campinas, Campinas-SP, Brazil Abstract: Natural products using plants have received considerable attention because of their potential to treat various diseases. Arrabidaea chica (Humb. & Bonpl. B. Verlot is a native tropical American vine with healing properties employed in folk medicine for wound healing, inflammation, and gastrointestinal colic. Applying nanotechnology to plant extracts has revealed an advantageous strategy for herbal drugs considering the numerous features that nanostructured systems offer, including solubility, bioavailability, and pharmacological activity enhancement. The present study reports the preparation and characterization of chitosan–sodium tripolyphosphate nanoparticles (NPs charged with A. chica standardized extract (AcE. Particle size and zeta potential were measured using a Zetasizer Nano ZS. The NP morphological characteristics were observed using scanning electron microscopy. Our studies indicated that the chitosan/sodium tripolyphosphate mass ratio of 5 and volume ratio of 10 were found to be the best condition to achieve the lowest NP sizes, with an average hydrodynamic diameter of 150±13 nm and a zeta potential of +45±2 mV. Particle size decreased with AcE addition (60±10.2 nm, suggesting an interaction between the extract’s composition

  9. Chica da Silva: Myth and Reality in an Extreme Case of Social Mobility

    Directory of Open Access Journals (Sweden)

    Maria Angélica Alves Pereira

    2014-06-01

    Full Text Available The story of Chica da Silva, a well-known historical figure in Brazil's popular culture, is examined, contrasting existing public records with myths about her life in Tijuco, the small town that became the world's center of diamond explotation in the XVIII century. Through her union with the King of Portugal's overseer of diamond extraction, this former slave gained access to a life of luxury and power far beyond that of other women of similar origins. Chica built a stable family, participated in religious organizations in her community, learned to write, and even supported artistic activities; while both written sources and many oral traditions depict her cruelty and promiscuity, these are contradicted by evidence of her social acceptance by the white elite and slaves alike. Myths can be best understood as diffuse but pervasive mechanisms of social control. Chica's trajectory remains a significant example of the power of individuals who believe in their own worth and ability to affect social change by altering expected patterns of superior/subordinate relationships.

  10. Provision of Oral Health Care to Children under Seven Covered by Bolsa Família Program. Is This a Reality?

    Science.gov (United States)

    Petrola, Krishna Andréia Feitosa; Bezerra, Ítalo Barroso; de Menezes, Érico Alexandro Vasconcelos; Calvasina, Paola; Saintrain, Maria Vieira de Lima; Pimentel G F Vieira-Meyer, Anya

    2016-01-01

    Over the last decade, there has been a great improvement in the oral health of Brazilians. However, such a trend was not observed among five-year-old children. Dental caries are determined by the interplay between biological and behavioral factors that are shaped by broader socioeconomic determinants. It is well established that dental disease is concentrated in socially disadvantaged populations. To reduce social and health inequalities, the Brazilian government created Family Health Program (ESF), and the Bolsa Família Program, the Brazilian conditional cash transfer program (Bolsa Família Program). The aim of this study was to examine the oral health care and promotion provided by the Family Health Teams to children and caregivers covered by the Bolsa Família Program. Data was collected through interviews with three groups of participants: 1) dentists working for the Family Health Program; 2) Family Health Program professionals supervising the Bolsa Família Program health conditionalities (Bolsa Família Program supervisors); and 3) parents/caregivers of children covered by the Bolsa Família Program. A pretested questionnaire included sociodemographic, Bolsa Família Program, oral health promotion, dental prevention and dental treatment questions. The results showed that most dentists performed no systematic efforts to promote oral health care to children covered by the Bolsa Família Program (93.3%; n = 69) or to their parents/caregivers (74.3%; n = 55). Many dentists (33.8%) did not provide oral health care to children covered by the Bolsa Família Program because they felt it was beyond their responsibilities. Nearly all Bolsa Família Program supervisors (97.3%; n = 72) supported the inclusion of oral health care in the health conditionality of the Bolsa Família Program, but 82.4% (n = 61) stated they did not promote oral health activities to children covered by the Bolsa Família Program. Children in the routine care setting were more often referred

  11. Manejo sostenible y sustentable de fincas productoras mediante procesos participativos en Sáchica, Boyacá

    OpenAIRE

    Ángel Eduardo Ramírez-Amaya; Germán Gonzalo Hurtado

    2013-01-01

    Objetivo. Elaborar un proyecto de desarrollo sostenible y sustentable de fincas productoras mediante procesos participativos en el municipio de Sáchica, Boyacá. Materiales y métodos. La investigación se realizó con familias campesinas de la vereda Arrayán Alto, del municipio de Sáchica, Boyacá, mediante la metodología Investigación Acción Participativa (IAP), que se centra en la participación de las comunidades para elaborar propuestas concertadas con ellas. El trabajo se desarrolló en varias...

  12. Autoimagem de clientes com colostomia em relação à bolsa coletora

    Directory of Open Access Journals (Sweden)

    Maria do Rosário de Fátima Franco Batista

    2011-12-01

    Full Text Available Objetivou-se analisar a percepção do portador de colostomia em relação ao uso da bolsa coletora. Realizou-se uma pesquisa descritiva com abordagem qualitativa, no Centro Integrado de Saúde Lineu Araújo, Teresina-PI. Participaram da pesquisa dez clientes portadores de bolsa de colostomia. Os dados foram produzidos por meio de entrevistas semiestruturadas. A análise de conteúdo permitiu revelar os sentimentos, as mudanças ocorridas e como acontece o processo de adaptação da pessoa portadora da bolsa de colostomia. Constatou-se que a relação entre a pessoa portadora de colostomia e a bolsa coletora é permeada por sentimentos negativos, mudanças significativas de ordem físicas, psicológicas, sexuais, bem como na teia de suas relações sociais.

  13. Ressecção de bolsa hiperfuncionante para tratamento de hipotonia ocular crônica: relato de casos

    OpenAIRE

    Cronemberger,Sebastião; Santos,Daniel Vítor de Vasconcelos; Oliveira,Ana Cláudia Monteiro; Maestrini,Heloísa Andrade; Calixto,Nassim

    2004-01-01

    Relatar os resultados obtidos com a ressecção de bolsa hiperfuncionante pós-trabeculectomia (TREC) com mitomicina C (MMC) para o tratamento da hipotonia ocular crônica. Cinco pacientes portadores de hipotonia ocular crônica causada por hiperfunção de bolsa fistulante pós- trabeculectomia com mitomicina foram tratados pela ressecção da bolsa. O diagnóstico de hiperfunção da bolsa foi feito com base em critérios estabelecidos pelos autores. A hipotonia ocular foi revertida nos cinco pacientes, ...

  14. Bolsas coletoras utilizadas por portadores de estoma: uma análise tridimensional

    Directory of Open Access Journals (Sweden)

    Jessica Andressa Collet

    2016-08-01

    Full Text Available No processo de cura de muitas doenças do intestino, diversas vezes, um procedimento cirúrgico conhecido por estomia é o único meio encontrado para manter o paciente em vida. Este procedimento se dá através da criação de uma abertura artificial no organismo, por onde acontece a saída das eliminações naturais do corpo, o levando, por este motivo, a utilizar uma bolsa externa para a coleta dos resíduos. O uso dessa bolsa coletora, para o estomizado, engloba uma série de questões físicas e psicológicas, que vão desde simples cuidados com o estoma até mesmo a incapacidade de retornar à vida social. A falta de informação e a utilização de dispositivos coletores de má qualidade expõem o estomizado a desconfortos, consistindo em causas frequentes para o seu isolamento. Sendo assim, o presente estudo consiste em apresentar a análise de bolsas coletoras, visando a verificação de aspectos (positivos ou negativos do aparelho. Para alcançar o objetivo, além da revisão teórica acerca do indivíduo estomizado e da bolsa coletora, foi utilizada a tecnologia da digitalização tridimensional por fotogrametria para que se pudesse obter uma maior realidade na análise das bolsas coletoras, uma vez que apenas imageticamente não se teria a verificação da realidade do estomizado. O uso da referida tecnologia possibilitou a análise de duas bolsas coletoras, fotografadas com volumes e poses variadas, cujos resultados permitiram a compreensão e comparação acerca do seu aspecto visual, segurança e discrição, onde se percebeu que a utilização desses produtos poderia auxiliar o estomizado no processo pós-cirúrgico e contribuir numa melhora na qualidade de vida. Esta pesquisa disponibiliza resultados que podem ser utilizados pela indústria e pela academia, instigando a possibilidade de trabalho com um nicho que carece de projetos de design.

  15. Bolsa Escola: Breaking the Cycle of Poverty, Child Labour and School Disaffection in Brazil

    Science.gov (United States)

    Denes, Christian Andrew

    2004-01-01

    The Bolsa Escola program in Brazil presents a clear break from the economic growth models and supply-side based strategies of the past. Founded on the assumption that the supplemental income generated by child labour outweighs the potential benefits of primary education, Bolsa Escola attempts to address the demand-side component of high dropout…

  16. Evaluation of wound healing properties of Arrabidaea chica Verlot extract.

    Science.gov (United States)

    Jorge, Michelle Pedroza; Madjarof, Cristiana; Gois Ruiz, Ana Lúcia Tasca; Fernandes, Alik Teixeira; Ferreira Rodrigues, Rodney Alexandre; de Oliveira Sousa, Ilza Maria; Foglio, Mary Ann; de Carvalho, João Ernesto

    2008-08-13

    Arrabidaea chica Verlot. (Bignoniaceae), popularly known as Crajiru, has been traditionally used as wound healing agent. Investigate in vitro and in vivo healing properties of Arrabidaea chica leaves extract (AC). AC was evaluated in vitro in fibroblast growth stimulation (0.25-250 microg/mL) and collagen production stimulation (250 microg/mL) assays. Allantoin (0.25-250 microg/mL) and vitamin C (25 microg/mL) were used as controls respectively. DPPH and Folin-Ciocalteau assays were used for antioxidant evaluation, using trolox (0.25-250 microg/mL) as reference antioxidant. To study wound healing properties in rats, AC (100mg/mL, 200 microL/wound/day) was topically administered during 10 days and wound area was evaluated every day. Allantoin (100mg/mL, 200 microL/wound/day) was used as standard drug. After treatment, wound sites were removed for histopathological analysis and total collagen determination. AC stimulated fibroblast growth in a concentration dependent way (EC50=30 microg/mL), increased in vitro collagen production and demonstrated moderate antioxidant capacity. In vivo, AC reduced wound size in 96%, whereas saline group showed only 36% wound healing. AC efficiency seems to involve fibroblast growing stimulus and collagen synthesis both in vitro and in vivo, beyond moderate scavenging activity, corroborating Crajiru folk use.

  17. Evolución de los pacientes con complicaciones locales en la bolsa del generador de un dispositivo implantable

    OpenAIRE

    Izquierdo,Maite; Bonanad,Clara; Madrazo,Inés; Ferrero,Ángel; Martínez,Ángel; Morell,Salvador; Chorro,Javier; Ruiz-Granell,Ricardo

    2014-01-01

    Objetivo: Las recomendaciones para la extracción completa de la bolsa de dispositivos implantables por problemas locales han cambiado. Analizamos la evolución entre 2002 y 2010 de los pacientes que requirieron una intervención por una complicación local en nuestro centro. Métodos: Ochenta y tres pacientes tuvieron un problema local de la bolsa que se clasificó según integridad de la piel: 1. Íntegra y 2. Abierta, y el tipo de intervención realizada: 1. Conservadora, 2. Extracción parcial y 3....

  18. Bolsa Família e desigualdade da renda domiciliar entre 2006 e 2011 = Bolsa Família and inequality of household income between 2006 and 2011

    Directory of Open Access Journals (Sweden)

    Carvalho, Cleusení Hermelina de

    2014-01-01

    Full Text Available Os programas de transferência condicionada de renda têm crescentemente desempenhado um papel importante no combate à pobreza em vários países da América Latina, principalmente no Brasil. O objetivo deste artigo é analisar a contribuição do programa Bolsa Família na diminuição da desigualdade da renda domiciliar per capita no Brasil, entre 2006 e 2011. Para isso, analisa-se a participação relativa de oito fontes de renda – trabalho, aposentadorias, programa Bolsa Família (variável proxy, pensões, abonos, doações, aluguéis e juros – no Brasil e nas suas cinco macrorregiões. Assim, além do artigo detalhar a técnica matemática utilizada para decompor o Índice de Gini, apresenta e discute os resultados empíricos encontrados para o Brasil e suas macrorregiões. Dentre os resultados, destaca-se a capacidade do programa Bolsa Família em contribuir para a queda da desigualdade da renda domiciliar nacional, o que se explica por sua acentuada focalização

  19. [Quality of food: perceptions of 'Bolsa Familia' program participants].

    Science.gov (United States)

    Uchimura, Kátia Yumi; Bosi, Maria Lúcia Magalhães; Lima, Flávia Emília Leite de; Dobrykopf, Vanessa França

    2012-03-01

    This study deals with perceptions of beneficiaries of the 'Bolsa Familia' Program, in Curitiba, southern Brazil, about their feeding habits. To understand the perceptions of participants of the 'Bolsa Família' Program on the quality of their food. A qualitative study based on the critical-interpretive tradition, which used individual interviews as a technique for gathering empirical data from the informants. The study included 38 individuals, members of families included in the program. The discursive content was recorded on digital media and, thereafter, transcribed and analyzed. After categorization, three main themes emerged: a description of food, quality of food, and feelings and experiences of individuals enrolled in the program. the acknowledgement of social vulnerability and consequent feeding habit insecurity to which such groups are subject was the main finding, as well as feelings of resignation.

  20. Exploration of agro-ecological options for improving maize-based farming systems in Costa Chica, Guerrero, Mexico

    NARCIS (Netherlands)

    Flores Sanchez, D.

    2013-01-01

    Keywords: farm diagnosis, farming systems, soil degradation, intercropping, maize, roselle, legumes, nutrient management, vermicompost, crop residues, decomposition, explorations.

    In the Costa Chica, a region of Southwest Mexico, farming systems are organized in

  1. Percepções de gênero entre casais beneficiários do Programa Bolsa Família

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    Mani Tebet

    2012-04-01

    Full Text Available Este artigo pretende perceber em que medida o Programa Bolsa Família modifica as relações de gênero, poder e interesse entre os casais beneficiários, tema que tem despertado pouco interesse do debate público e acadêmico. Pretende ainda sinalizar ecompreender os efeitos morais e simbólicos que a política pode produzir sobre a família e as relações de gênero; os critérios de justiça apontados pelos casais para “merecer” o Bolsa Família; e a lógica que se encontra na base dessa noção de merecimento. Paratanto, entrevistamos casais beneficiários do bairro de Nova Cidade, no município de Itaboraí, na Região Metropolitana do Rio de Janeiro. Gender Perceptions between Couples Benefiting from the ‘Bolsa Família’ Program intends to ascertain to what extent the ‘Bolsa Família’ allowance program modifies the gender, power and interest relations between couples who receive benefits from the scheme; a topic that has aroused very little interest in public and academic debate. It also attempts to identify and understand the moral and symbolic effects that the policy can produce on the family and gender relations; the criteria of justice put forward by the couples for “deserving” the Bolsa Família; and the rationale that underlies this notion of deserving. Interviews were conducted with couples in a district in the metropolitan region of Rio de Janeiro.Keywords: poverty, gender, family, ‘Bolsa Família’ Allowance Program, income transfer

  2. Regras importam: determinantes do controle burocrático no Programa Bolsa Família

    OpenAIRE

    Coêlho, Denilson Bandeira; Fernandes, Antônio Sérgio Araújo

    2017-01-01

    Resumo: A literatura concernente ao Programa Bolsa Família tem focado questões como o impacto sobre a pobreza e a desigualdade, os efeitos relacionados com o processo eleitoral e a função das condicionalidades. Entretanto, o Programa Bolsa Família está também associado a um problema de relação principal-agente, pois requer o controle efetivo de um conjunto de regras para seu funcionamento. Na literatura nacional, pouco se tem produzido sobre o efeito de regras formais como instrumentos de mod...

  3. Environmental evidence of fossil fuel pollution in Laguna Chica de San Pedro lake sediments (Central Chile)

    International Nuclear Information System (INIS)

    Chirinos, L.; Rose, N.L.; Urrutia, R.; Munoz, P.; Torrejon, F.; Torres, L.; Cruces, F.; Araneda, A.; Zaror, C.

    2006-01-01

    This paper describes lake sediment spheroidal carbonaceous particle (SCP) profiles from Laguna Chica San Pedro, located in the Biobio Region, Chile (36 o 51' S, 73 o 05' W). The earliest presence of SCPs was found at 16 cm depth, corresponding to the 1915-1937 period, at the very onset of industrial activities in the study area. No SCPs were found at lower depths. SCP concentrations in Laguna Chica San Pedro lake sediments were directly related to local industrial activities. Moreover, no SCPs were found in Galletue lake (38 o 41' S, 71 o 17.5' W), a pristine high mountain water body used here as a reference site, suggesting that contribution from long distance atmospheric transport could be neglected, unlike published data from remote Northern Hemisphere lakes. These results are the first SCP sediment profiles from Chile, showing a direct relationship with fossil fuel consumption in the region. Cores were dated using the 21 Pb technique. - The lake sediment record of SCPs shows the record of fossil-fuel derived pollution in Central Chile

  4. Environmental evidence of fossil fuel pollution in Laguna Chica de San Pedro lake sediments (Central Chile)

    Energy Technology Data Exchange (ETDEWEB)

    Chirinos, L. [Centro de Ciencias Ambientales EULA-Chile, Universidad de Concepcion, PO Box 160-C, Concepcion (Chile)]. E-mail: lchirin@pucp.edu.pe; Rose, N.L. [Environmental Change Research Centre, University College London, 26 Bedford Way, London WG1HOAP (United Kingdom); Urrutia, R. [Centro de Ciencias Ambientales EULA-Chile, Universidad de Concepcion, PO Box 160-C, Concepcion (Chile); Munoz, P. [Departamento de Biologia Marina, Universidad Catolica del Norte, Larrondo 1281, Coquimbo (Chile); Torrejon, F. [Centro de Ciencias Ambientales EULA-Chile, Universidad de Concepcion, PO Box 160-C, Concepcion (Chile); Torres, L. [Departamento de Botanica, Universidad de Concepcion, Concepcion (Chile); Cruces, F. [Departamento de Botanica, Universidad de Concepcion, Concepcion (Chile); Araneda, A. [Centro de Ciencias Ambientales EULA-Chile, Universidad de Concepcion, PO Box 160-C, Concepcion (Chile); Zaror, C. [Facultad de Ingenieria Quimica, Universidad de Concepcion, Concepcion (Chile)

    2006-05-15

    This paper describes lake sediment spheroidal carbonaceous particle (SCP) profiles from Laguna Chica San Pedro, located in the Biobio Region, Chile (36{sup o} 51' S, 73{sup o} 05' W). The earliest presence of SCPs was found at 16 cm depth, corresponding to the 1915-1937 period, at the very onset of industrial activities in the study area. No SCPs were found at lower depths. SCP concentrations in Laguna Chica San Pedro lake sediments were directly related to local industrial activities. Moreover, no SCPs were found in Galletue lake (38{sup o} 41' S, 71{sup o} 17.5' W), a pristine high mountain water body used here as a reference site, suggesting that contribution from long distance atmospheric transport could be neglected, unlike published data from remote Northern Hemisphere lakes. These results are the first SCP sediment profiles from Chile, showing a direct relationship with fossil fuel consumption in the region. Cores were dated using the {sup 21}Pb technique. - The lake sediment record of SCPs shows the record of fossil-fuel derived pollution in Central Chile.

  5. Influence of the Bolsa Família program on nutritional status and food frequency of schoolchildren.

    Science.gov (United States)

    do Carmo, Ariene Silva; de Almeida, Lorena Magalhães; de Oliveira, Daniela Rodrigues; Dos Santos, Luana Caroline

    2016-01-01

    To evaluate the food frequency and nutritional status among students according to participation in the Bolsa Família program funded by the government. Cross-sectional study carried out with students from the fourth grade of elementary school in the municipal capital of the southeastern region of Brazil. Food consumption and anthropometry were investigated by a questionnaire administered in school, while participation in the Bolsa Família program and other socio-economic information was obtained through a protocol applied to mothers/guardians. Statistical analysis included the Mann-Whitney test, the chi-squared test, and Poisson regression with robust variance, and the 5% significance level was adopted. There were 319 children evaluated; 56.4% were male, with a median of 9.4 (8.6-11.9) years, and 37.0% were beneficiaries of Bolsa Família program. Between the two groups, there was high prevalence of regular soda consumption (34.3%), artificial juice (49.5%), and sweets (40.3%), while only 54.3% and 51.7% consumed fruits and vegetables regularly, respectively. Among participants of Bolsa Família program, a prevalence 1.24 times higher in the regular consumption of soft drinks (95% CI: 1.10-1.39) was identified compared to non-beneficiaries. The prevalence of overweight was higher in the sample (32.9%), with no difference according to participation in the program. The study found increased consumption of soft drinks among BFP participants. The high rate of overweight and poor eating habits denote the need to develop actions to promote healthy eating, especially for the beneficiaries of the Bolsa Família program, to promote improvements in nutritional status and prevent chronic diseases throughout life. Copyright © 2016 Sociedade Brasileira de Pediatria. Published by Elsevier Editora Ltda. All rights reserved.

  6. Manejo sostenible y sustentable de fincas productoras mediante procesos participativos en Sáchica, Boyacá

    Directory of Open Access Journals (Sweden)

    Ángel Eduardo Ramírez-Amaya

    2013-07-01

    Full Text Available Objetivo. Elaborar un proyecto de desarrollo sostenible y sustentable de fincas productoras mediante procesos participativos en el municipio de Sáchica, Boyacá. Materiales y métodos. La investigación se realizó con familias campesinas de la vereda Arrayán Alto, del municipio de Sáchica, Boyacá, mediante la metodología Investigación Acción Participativa (IAP, que se centra en la participación de las comunidades para elaborar propuestas concertadas con ellas. El trabajo se desarrolló en varias fases, que incluyeron un diagnóstico socioeconómico de las familias, capacitaciones y concientización en temas relacionados con la agricultura ecológica y de granjas integrales. Resultados. Se elaboró un plan de trabajo que permitió la construcción de un documento final que ha servido para el apoyo logístico o económico de las entidades gubernamentales locales para la instalación y plantación técnica del cultivo de gulupa con familias de la vereda Arrayán Alto.

  7. Motivações para comprar objeto de luxo: Bolsas LV

    Directory of Open Access Journals (Sweden)

    Luis Alexandre Grubits de Paula Pessôa

    2012-10-01

    Full Text Available Este estudo exploratório investigou motivações que levam mulheres de classe média a adquirir um objeto de luxo, apesar do impacto da compra em seu orçamento. Baseado no modelo das cadeias meios-fim, conduziram-se entrevistas com 15 mulheres que adquiriram algum modelo genuíno de bolsa Louis Vuitton. Os resultados sugerem que os grupos de referência exercem tanto influência normativa quanto de identificação na decisão de compra. Mais importante do que os atributos do acessório declarados, as mulheres entrevistadas consideram que a posse e a ostentação da bolsa conferem a elas status e prestígio em seus grupos de referência, criando também a sensação de serem aceitas em grupos de aspiração, elevando a autoestima e levando-as a se perceberem profissionalmente bem sucedidas, valores individuais que parecem ser os reais motivadores da decisão de compra.

  8. Ansiedad física social y educación física escolar: las chicas adolescentes en las clases de natación

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    MarÍa José Camac ho-Miñano

    2014-06-01

    Full Text Available Este estudio cualitativo analiza la ansiedad física social (AFS que experimentan las chicas adolescentes en el contexto de las clases de natación que se desarrollan en educación física (EF, profundizando en sus factores explicativos así como en las estrategias de afrontamiento (coping utilizadas. Para ello se ha realizado un estudio de caso en un centro escolar mediante 12 entrevistas semiestructuradas a chicas adolescentes y al profesorado que les imparte clase de natación junto con la observación de las clases. Los datos se han categorizado mediante un análisis de contenido cualitativo y triangulación. Los resultados señalan que la AFS es una emoción que muchas chicas experimentan en las clases de natación debido a que el cuerpo queda expuesto a las miradas evaluativas de los demás, especialmente de los chicos. El contexto social de las clases se revela como clave ya que sufrir críticas o burlas en relación a la apariencia física por parte de los compañeros/as es uno de los factores causales de la AFS, mientras que el apoyo y aceptación social contribuyen a minimizarla. Las consecuencias de este malestar derivan en estrategias de coping orientadas a: la resolución de problemas adoptando conductas específicas para ocultar el cuerpo; el manejo de las emociones, autoconvenciéndose de que es una situación normal; e incluso, la evitación de la situación negándose a participar en las clases. Para proporcionar a las chicas una experiencia positiva de la EF, el profesorado debería considerar esta problemática en la enseñanza de este contenido.

  9. Exploration of agro-ecological options for improving maize-based farming systems in Costa Chica, Guerrero, Mexico

    OpenAIRE

    Flores Sanchez, D.

    2013-01-01

    Keywords: farm diagnosis, farming systems, soil degradation, intercropping, maize, roselle, legumes, nutrient management, vermicompost, crop residues, decomposition, explorations. In the Costa Chica, a region of Southwest Mexico, farming systems are organized in smallholder units. The dominant cropping systems are based on maize (Zea mays L.), either as monocrop or intercropped with roselle (Hibiscus sabdariffa L.). Continuous cropping, and unbalanced fertilizer management systems with an...

  10. Redes Neuronales y su aplicación predictiva en la Bolsa de Valores española

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    Ibarra Alfaraz, J.A.

    1999-01-01

    Full Text Available Este trabajo recoge una visión general de las redes neuronales y su tendencia en la investigación actual. Además de las aplicaciones conocidas donde las redes neuronales han demostrado su aplicabilidad se abren nuevos campos de investigación. En nuestro caso, nos centraremos en el área económica y más concretamente en la financiera. Esta metodología se aplica al caso concreto del análisis predictivo de la bolsa de valores, concretamente se utiliza el índice del mercado continuo de la bolsa española, Ibex-35, y los recientemente aparecidos índices sectoriales del Ibex: Servicios, Financiero, Utilities y Complementario. La red neuronal utilizada, Perceptron, ha sido entrenada con los datos reales procedentes de la bolsa de valores utilizando diferentes periodos de tiempo y efectuando cambios en los parámetros que condicionan la capacidad predictiva de la red. Los resultado se han contrastado con los obtenidos en otros trabajos empíricos realizados con metodologías clásicas.

  11. A produção da maternidade no Programa Bolsa-Escola The production of maternity in the Bolsa-Escola Program

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    Carin Klein

    2005-04-01

    Full Text Available Neste trabalho, analiso alguns processos de produção e veiculação de representações de maternidade, tomando como referência o Programa Nacional Bolsa-Escola. Meu estudo insere-se nos campos dos Estudos Culturais e dos Estudos Feministas, nas vertentes que têm proposto uma aproximação crítica com a abordagem pós-estruturalista. Para a operacionalização da pesquisa, selecionei um conjunto de documentos referentes a esse Programa, produzidos e publicados no período de 1999 a 2003. Exploro os textos do Programa tomando como base os conceitos de discurso, representação, identidade, gênero e poder com o intuito de analisar os diferentes modos pelos quais a maternidade é, ali, representada e significada.In this work I analyze some processes of production and conveyance of maternity representations, having the Programa Nacional Bolsa-Escola as its reference. My study is located in the field of cultural theory, mainly in the Cultural Studies and Feminist Studies perspectives, in approaches that have proposed a critical approximation to the post-structuralist analysis. In order to perform this research, I selected a set of documents related to the Program, which were produced and published from 1999 to 2003. I have explored the texts of the Program on the basis of concepts such as discourse, representation, identity, gender and power, aiming at analyzing the different ways by which maternity has been represented and signified there.

  12. Embolia gasosa venosa inadvertida durante cesariana: bolsas retráteis ​​para líquidos intravenosos sem saídas autovedantes oferecem riscos. Relato de caso

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    Mefkur Bakan

    2013-08-01

    Full Text Available O anestesiologista deve estar ciente das causas, do diagnóstico e do tratamento de embolia venosa e adotar padrões de prática para prevenir sua ocorrência. Embora a embolia gasosa seja uma complicação conhecida da cesariana, descrevemos um caso raro de desatenção que causou embolia gasosa iatrogênica quase fatal durante uma cesariana sob raquianestesia. uma das razões para o uso de bolsas autorretráteis para infusão em vez dos frascos convencionais de vidro ou plástico é a precaução contra embolia gasosa. Também demonstramos o risco de embolia venosa com o uso de dois tipos de bolsas plásticas retráteis (à base de cloreto de polivinil [PVC] e de polipropileno para líquidos intravenosos. As bolsas para líquidos sem saídas autovedantes apresentam risco de embolia gasosa se o sistema de fechamento estiver quebrado, enquanto a flexibilidade da bolsa limita a quantidade de entrada de ar. bolsas à base de pvc, que têm mais flexibilidade, apresentam risco significativamente menor de entrada de ar quando o equipo de administração intravenosa (IV é desconectado da saída. usar uma bolsa pressurizada para infusão rápida sem verificar e esvaziar todo o ar da bolsa IV pode ser perigoso.

  13. EVALUACIÓN DE LA CALIDAD MICROBIOLÓGICA DEL AGUA ENVASADA EN BOLSAS PRODUCIDA EN SINCELEJO- COLOMBIA

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    Jhon Vidal D

    2009-08-01

    Full Text Available Objetivo. Evaluar la calidad microbiológica y fisico-quimica del agua envasada en bolsas producidad en la ciudad de Sincelejo-Colombia con destino al consumo humano. Materiales y métodos. Para la estimación de organismos coliformes totales y fecales, Pseudomona aeruginosa y mesófilos en el agua envasada de 13 marcas, se utilizó el método de filtración por membrana (FxM. Resultados. El 92 % de las marcas de agua envasada en bolsa que se produce en la ciudad de Sincelejo presentaron bacterias mesófilas en su producto, mientras que en el 33% de ellas se encontraron coliformes totales. Cabe destacar que una marca presentó coliformes fecales, otra Pseudomonas aeruginosa y el reporte microbiano fue mayor en las envasadoras que poseían registro INVIMA. Conclusiones. Gran parte del agua envasada en bolsas de la ciudad de Sincelejo genera un riesgo a la salud de los consumidores, debido a la presencia de microorganismos patógenos, lo que está relacionado con inadecuados procesos de producción y a la intermitencia del suministro del agua utilizada como materia prima.

  14. La Bolsa de Valores de México durante el porfiriato y la revolución, 1885-1934

    OpenAIRE

    Javier Moreno-Lázaro

    2017-01-01

    En este artículo se sostiene que la Bolsa de México sólo contribuyó a la financiación empresarial desde su fundación hasta el inicio de la revolución. Allí encontraron fuentes de financiación empresarios mineros, banqueros e industriales. Pero desde entonces, particularmente desde la aplicación de la doctrina de Carranza en 1916, la Bolsa se convirtió en un mero instrumento financiero del Estado y sirvió entonces casi exclusivamente para la suscripción de deuda. Para demostrar esta hipótesis ...

  15. Asociaciones políticas de inmigrantes peruanos y la "Lima Chica" en Santiago de Chile

    OpenAIRE

    Luque Brazán, José Carlos

    2007-01-01

    El presente trabajo describe y examina la emergencia y desarrollo de tres asociaciones políticas de inmigrantes peruanos y su relación con el surgimiento de un "vecindario cultural", conocido por sus habitantes, la prensa chilena y algunos investigadores como la "Lima Chica", en Santiago de Chile. Nos referimos al Comité de Refugiados Peruanos en Chile, a la Asociación de Inmigrantes por la Integración Latinoamericana y del Caribe (APILA) y al Programa Andino para la Dignidad Humana (Proandes...

  16. Prácticas sexuales de chicos y chicas españoles de 14-24 años de edad Sexual behavior in a Spanish sample aged 14 to 24 years old

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    José María Faílde Garrido

    2008-12-01

    Full Text Available Objetivo: Describir los comportamientos y prácticas sexuales de adolescentes y jóvenes españoles en función del género. Método: La información fue recogida mediante un cuestionario, realizado en el domicilio de los participantes y con presencia del entrevistador, aplicado a una muestra aleatoria integrada por 2.171 chicos y chicas de 14-24 años de edad, representativa de las comunidades de Galicia, Madrid y Andalucía. Resultados: Un total de 1.439 sujetos (66,3% refirieron haber tenido actividad sexual en los últimos 6 meses, sin apreciarse diferencias estadísticamente significativas entre chicos (66,4% y chicas (66,2%, excepto en las siguientes variables: haber practicado el coito anal (los chicos refieren haberlo practicado en mayor proporción; número de parejas sexuales (las chicas manifestaron tener menor número de parejas, y frecuencia de coitos vaginales (las chicas presentaron una frecuencia más elevada en esta práctica. También se encontraron diferencias en frecuencia de uso del condón en las prácticas coito-anales y en las bucogenitales, en las que los chicos refirieron utilizarlo más frecuentemente. Conclusiones: Los datos de este estudio indican que los chicos y las chicas mantienen comportamientos sexuales diferenciados. En este sentido, las chicas suelen tener menor número de parejas sexuales y utilizan el preservativo en mayor medida que los chicos en las prácticas coito-vaginales; sin embargo, hacen menor uso de éste en las prácticas bucogenitales y coito-anales. En función de estos datos consideramos necesario tener en cuenta la variable género a la hora de diseñar e implementar intervenciones preventivas.Objectives: To describe the sexual behaviors and practices of Spanish adolescents and young adults according to gender. Method: Information was gathered by means of a questionnaire administered in participants' homes in the presence of an interviewer. A random sample was used, consisting of 2

  17. Society-State relationships, citizen participation and political clientelism inside programs that combat poverty. The case of «Bolsa Familia» in Brazil

    Directory of Open Access Journals (Sweden)

    Felipe J. HEVIA

    2011-06-01

    Full Text Available Relations between poor people and the government that creates the Programa Bolsa Familia at Brazil may be summarized in two dimensions: 1 favor direct relationships without the intervention of collective action and 2 are distant relations in terms of type of interaction and communication between the authorities and beneficiaries. While there are instances of formal social control, operation of the program makes minimal intermediation and highly institutional and civic organizations have little room to act and to represent the beneficiaries of Bolsa Familia in institutionalized interfaces. Direct links generate positive effect low levels of political patronage vote buying and coercion, but also generate unintended effects such as the lack of program operation, difficulty to defend themselves collectively by irregularities and create an active citizenry.

  18. Impact of the Bolsa Família program on food availability of low-income Brazilian families: a quasi experimental study.

    Science.gov (United States)

    Martins, Ana Paula Bortoletto; Monteiro, Carlos Augusto

    2016-08-19

    The Bolsa Família Program was created in Brazil in 2003, by the joint of different social programs aimed at poor or very poor families with focus on income transfer to promote immediate poverty relief, conditionalities and complementary programs. Given the contributions of conditional cash transfer programs to poverty alleviation and their potential effects on nutrition and health, the objective of this study was to assess the impact of the Bolsa Família Program on food purchases of low-income households in Brazil. Representative data from the Household Budget Survey conducted in 2008-2009 were studied, with probabilistic sample of 55,970 households. 11,282 households were eligible for this study and 48.5 % were beneficiaries of the BFP. Food availability indicators were compared among paired blocks of households (n = 100), beneficiaries or non-beneficiaries of the Bolsa Família Program, with monthly per capita income up to R$ 210.00. Blocks of households were created based on the propensity score of each household to have beneficiaries and were homogeneous regarding potential confounding variables. The food availability indicators were weekly per capita expenditure and daily energy consumption, both calculated considering all food items and four food groups based on the extent and purpose of the industrial food processing. The comparisons between the beneficiaries and non-beneficiaries blocks of households were conducted through paired 't' tests. Compared to non-beneficiaries, the beneficiaries households had 6 % higher food expenditure (p = 0.015) and 9.4 % higher total energy availability (p = 0.010). It was found a 7.3 % higher expenditure on in natura or minimally processed foods and 10.4 % higher expenditure on culinary ingredients among the Bolsa Família Program families. No statistically significant differences were found regarding the expenditure and the availability of processed and ultra-processed food and drink products. In the in

  19. Bolsa Família (Family Grant) Programme: an analysis of Brazilian income transfer programme

    NARCIS (Netherlands)

    L. Mourao (Luciana); A. Macedo de Jesus (Anderson)

    2012-01-01

    markdownabstract__Abstract__ Income transfer programmes are common in various countries and play an important role in combating poverty. This article presents a review of the results of the Bolsa Família (Family Grant) Programme, implemented in Brazil by the government of Lula da Silva in

  20. Atividade de extratos de Arrabidaea chica (Humb. & Bonpl. Verlot obtidos por processos biotecnológicos sobre a proliferação de fibroblastos e células tumorais humanas

    Directory of Open Access Journals (Sweden)

    Denise Taffarello

    2013-01-01

    Full Text Available Arrabidaea chica (H&B Verlot is a plant popularly known as Pariri and this species is a known source of anthocyanins, flavonoids and tannins. This report describes an approach involving enzymatic treatment prior to extraction procedures to enhance A chica crude extract anticancer activity. Anticancer activity in human cancer cell lines in vitro using a 48 h SRB cell viability assay was performed to determine growth inhibition and cytotoxic properties. The final extraction yield without enzyme treatment was higher (24.28% compared to the enzyme-treated material (19.03%, with an enhanced aglycones anthocyanin ratio as determined by HPLC- DAD and LC-MS with direct infusion.

  1. Avaliação de Impacto das condicionalidades de educação do Programa Bolsa Família (2005 e 2009

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    Ernesto Friedrich de Lima Amaral

    2013-09-01

    Full Text Available Dans cet article, on examine les impacts des conditionnalités de l'éducation dans le Programme Bolsa Família sur l'absentéisme scolaire d'enfants qui bénéficient de ce programme. L'hypothèse principale est que l'enfant qui habite dans un foyer recevant cette aide a moins de chances d'abandonner l'école. On se sert de données de l'Étude de l'impact du Programme Bolsa Família (AIBF de 2005 à 2009 du Ministère du Développement Social et de la Lutte contre la Faim (MDS. Des modèles logistiques ont estimé les chances d'abandon scolaire de 2005 à 2009, à partir de trois niveaux de revenu domiciliaire par habitant, compte tenu des caractéristiques du foyer, de la mère et de l'enfant. Les enfants habitant dans des foyers bénéficiaires du Programme Bolsa Família ont révélé une nette réduction du taux d'abandon scolaire en 2005. Les données pour 2009 n'ont pas été statistiquement significatives, bien que montrant une diminution de l'abandon scolaire comme résultat de l'aide reçue du Bolsa Família.

  2. Integração do enxerto heterólogo de pele humana no subepitélio da bolsa jugal do hamster (Mesocricetus auratus

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    Hochman Bernardo

    2003-01-01

    Full Text Available OBJETIVO: Descrever a integração de enxertos de pele total humana no subepitélio da bolsa jugal do hamster (Mesocricetus auratus. MÉTODOS: A amostragem consistiu de 18 hamsters machos, exogâmicos, com 10 a 14 semanas de idade. Fragmentos de pele humana normal foram obtidos de pele excedente de mastoplastia redutora de paciente parda. Cada hamster foi enxertado em ambas as bolsas com fragmentos de pele, perfazendo um total de 36 fragmentos enxertados. Os animais foram distribuídos, em 6 grupos, para exame dos fragmentos enxertados com 5, 12, 21, 42, 84 e 168 dias. Uma avaliação macroscópica foi realizada comparando a bolsa contendo o fragmento enxertado em cada período com a mesma bolsa no pós-operatório imediato, mediante fotografias padronizadas. Na avaliação microscópica foi adotado como critério de integração a presença de vasos sangüíneos na derme dos enxertos. Observou-se também a presença de queratina, melanócitos, infiltrado celular e aspecto do tecido conjuntivo. RESULTADOS: Na avaliação macroscópica foi observada uma reação vascular em torno dos fragmentos até 12 dias do implante, e a presença de pigmentação castanho-escura a partir de 42 dias. À microscopia, integraram-se 80,64% dos fragmentos enxertados, inclusive no grupo de 168 dias. Observou-se infiltrado celular inflamatório até 12 dias, a presença de melanócitos a partir de 42 dias e uma hialinização do tecido conjuntivo após 84 dias. CONCLUSÕES: Fragmentos de pele humana integram-se no tecido celular subcutâneo da bolsa jugal do hamster, mantêm-se vascularizados por 168 dias, e conservam o epitélio íntegro até 21 dias. O subepitélio da bolsa representa modelo experimental de investigação da fisiologia de pele humana ex vivo.

  3. O PROGRAMA BOLSA FAMÍLIA E O “EFEITO PREGUIÇA”

    Directory of Open Access Journals (Sweden)

    Maria Luiza Souza Caetano

    2016-07-01

    Full Text Available During the implementation of the Bolsa Família Program emerged some myths like "laziness effect". This present study aims to reflect on the existence of laziness effect in terms of the works of Campello and Neri (2013 and Weissheimer (2006. Methodologically it is a theoretical article. For this, was made a review of the program, cash transfers and laziness effect. Finally, from the reflections on the works of the authors, it is clear that the laziness effect is a myth.

  4. La Bolsa de Valores de México durante el porfiriato y la revolución, 1885-1934

    Directory of Open Access Journals (Sweden)

    Javier Moreno-Lázaro

    2017-01-01

    Full Text Available En este artículo se sostiene que la Bolsa de México sólo contribuyó a la financiación empresarial desde su fundación hasta el inicio de la revolución. Allí encontraron fuentes de financiación empresarios mineros, banqueros e industriales. Pero desde entonces, particularmente desde la aplicación de la doctrina de Carranza en 1916, la Bolsa se convirtió en un mero instrumento financiero del Estado y sirvió entonces casi exclusivamente para la suscripción de deuda. Para demostrar esta hipótesis se presentan series cuantitativas inéditas que miden los efectos de la actividad bursátil en el desarrollo económico del país, al margen de los aspectos específicamente financieros.

  5. O empoderamento feminino e as mulheres do programa Bolsa Família

    OpenAIRE

    Williams, Priscila

    2016-01-01

    A luta por empoderamento das mulheres remonta às primeiras lutas feministas, mas isso parece ser ainda mais difícil para as mulheres pobres. Neste trabalho, busca-se compreender o processo de empoderamento das mulheres bene ciárias do Programa Bolsa Família, a partir do início do recebimento do benefício. A partir do relato dessas mulheres, pode-se observar que ainda há muito que ser feito para o efetivo rompimento da pobreza. 

  6. Armazenamento de soja em silos tipo bolsa Soybean storage in bag type silos

    Directory of Open Access Journals (Sweden)

    Lêda R. A. Faroni

    2009-03-01

    Full Text Available Avaliaram-se as principais alterações qualitativas de soja armazenada em silos tipo bolsa e do óleo bruto extraído de soja com teores de água de 17,4% e 13,3%, armazenada em dois silos tipo bolsa, por 180 dias. Realizaram-se amostragens no dia do enchimento das bolsas, aos 30; 90 e 180 dias de armazenamento. Analisaram-se o teor de água, a condutividade elétrica, o percentual de germinação, a massa específica aparente da soja, além do teor de ácidos graxos livres e o índice de peróxido do óleo bruto extraído dela. Os teores de água da soja armazenada úmida e seca mantiveram-se próximos dos valores obtidos no início do período de armazenamento. Observou-se tendência de elevação da condutividade elétrica e decréscimo do percentual de germinação somente na soja úmida, principalmente após 90 dias de armazenamento. Não foi verificado decréscimo da massa específica aparente do material armazenado úmido e seco. Com relação aos parâmetros qualitativos do óleo bruto, observou-se que os valores obtidos se mantiveram abaixo do limite máximo exigido pela legislação para a comercialização de óleo bruto de soja. Pode-se concluir que os silos tipo bolsa representam alternativa viável do ponto de vista qualitativo para armazenagem de soja, e esse tipo de estrutura não ocasiona alterações qualitativas significativas no óleo bruto obtido desse material, em condições similares àquelas deste estudo.This study reports major qualitative changes in the soybean grains and the extracted crude oil when stored in bag type silos. Grains with moisture content of 17.4 or 13.3% were stored in two bag type silos. Samples were taken 30, 90 and 180 days of storage , to determine moisture content, electric conductivity of the grain leachate, germination percentage, apparent specific grain mass, and free fatty acid content, and peroxide index of the crude oil extracted from these grains. The wet and dry grains remained with

  7. Influence of the Bolsa Família program on nutritional status and food frequency of schoolchildren

    Directory of Open Access Journals (Sweden)

    Ariene Silva do Carmo

    2016-07-01

    Full Text Available Objective: To evaluate the food frequency and nutritional status among students according to participation in the Bolsa Família program funded by the government. Methods: Cross-sectional study carried out with students from the fourth grade of elementary school in the municipal capital of the southeastern region of Brazil. Food consumption and anthropometry were investigated by a questionnaire administered in school, while participation in the Bolsa Família program and other socio-economic information was obtained through a protocol applied to mothers/guardians. Statistical analysis included the Mann–Whitney test, the chi-squared test, and Poisson regression with robust variance, and the 5% significance level was adopted. Results: There were 319 children evaluated; 56.4% were male, with a median of 9.4 (8.6–11.9 years, and 37.0% were beneficiaries of Bolsa Família program. Between the two groups, there was high prevalence of regular soda consumption (34.3%, artificial juice (49.5%, and sweets (40.3%, while only 54.3% and 51.7% consumed fruits and vegetables regularly, respectively. Among participants of Bolsa Família program, a prevalence 1.24 times higher in the regular consumption of soft drinks (95% CI: 1.10–1.39 was identified compared to non-beneficiaries. The prevalence of overweight was higher in the sample (32.9%, with no difference according to participation in the program. Conclusion: The study found increased consumption of soft drinks among BFP participants. The high rate of overweight and poor eating habits denote the need to develop actions to promote healthy eating, especially for the beneficiaries of the Bolsa Família program, to promote improvements in nutritional status and prevent chronic diseases throughout life. Resumo: Objetivo: Avaliar a frequência alimentar e estado nutricional entre escolares segundo a participação no programa governamental Bolsa Família (PBF. Metodologia: Estudo de delineamento

  8. Efecto de la Irrigación Crevicular con Azitromicina y con Tetraciclina en el Periodonto de Revestimiento y de Soporte en Pacientes Sometidos a Curetaje de Bolsa en el Centro Odontológico Dentalplans Arequipa 2009

    OpenAIRE

    Gonzales Calderón Juan Carlos

    2010-01-01

    La presente investigación tuvo como propósito central determinar el efecto de la Irrigación crevicular con azitromicina y con tetraciclina en el Periodonto de Revestimiento y de soporte en pacientes sometidos a curetaje de bolsa. La Investigación es cuasi experimental emparejado (intrasujeto) prospectiva, longitudinal, comparativa y de campo. Se conformó un grupo de estudio dividido en 2 sectores experimentales, cada uno de los cuáles estuvo constituido por 31 bolsas peri...

  9. O Programa Institucional de Bolsas de Iniciação à Docência (PIBID e as relações público/privadas no ensino superior

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    Margarita Victoria Rodríguez

    2017-04-01

    Full Text Available O objeto deste trabalho é o Programa Institucional de Bolsas de Iniciação à Docência (PIBID, sendo seu objetivo analisar a regulamentação e a aprovação de bolsas e subprojetos do PIBID, sob a perspectiva da relação público/privada no ensino superior. Para tal, o procedimento metodológico utilizado foi a análise da legislação do Programa, bem como os dados disponibilizados nos relatórios do PIBID. Isto posto, é evidenciado que, embora pese o caráter neoliberal na organização do Programa, com relação à parceria público/privada, o maior volume de bolsas e projetos aprovados ainda estavam com o setor público, até 2013. Algumas tensões são apresentadas nessa perspectiva, mas elas também evidenciam a precarização do ensino superior privado e a característica assistencialista do programa.

  10. Influence of the Bolsa Família program on nutritional status and food frequency of schoolchildren

    Directory of Open Access Journals (Sweden)

    Ariene Silva do Carmo

    2016-07-01

    Conclusion: The study found increased consumption of soft drinks among BFP participants. The high rate of overweight and poor eating habits denote the need to develop actions to promote healthy eating, especially for the beneficiaries of the Bolsa Família program, to promote improvements in nutritional status and prevent chronic diseases throughout life.

  11. Consumo alimentar de beneficiários do programa Bolsa Família

    OpenAIRE

    Alan Giovanini de Oliveira Sartori

    2014-01-01

    A expansão do consumo de alimentos submetidos a elevado grau de processamento em países em desenvolvimento é notória. Em paralelo, observa-se o aumento na prevalência de excesso de peso e de comorbidades associadas. O fenômeno também tem sido observado em famílias consideradas pobres que recebem benefício financeiro de programa federal de transferência condicionada de renda. O objetivo geral foi analisar o consumo alimentar de beneficiários do Programa Bolsa Família (PBF). Foi elaborado um si...

  12. Programa bolsa família: a condicionante saúde realmente existe?

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    Núbia Maria Uchôa Barbosa

    2014-12-01

    Full Text Available Historicamente, o Sistema de Proteção Social do Brasil se caracteriza por apresentar uma estrutura dual de seguridade social: aos grupos mais vulneráveis socialmente e não inseridos no mercado de trabalho, destina-se a assistência social, enquanto os trabalhadores inseridos no mercado formal de trabalho vinculam-se à previdência social. As camadas pobres da sociedade brasileira, marcadas pela quase ausência de pressão social e sem posição sócio-ocupacional definida, em alguns momentos históricos, foram beneficiadas, e seu atendimento sempre foi justificado como um ato humanitário ou uma moeda política(1. Nesse aspecto, destaca-se o Programa Bolsa Família (PBF, como um programa de combate à pobreza, criado através de Medida Provisória n.o 132/2003, transformado em Lei n.o 10.836/2004 e regulamentado por Decreto n.o 5.209/2004. Foi iniciado em outubro de 2003 e constituído através da unificação de quatro programas de transferência de renda: Bolsa Escola, Auxílio-Gás, Bolsa Alimentação e Cartão Alimentação(2. A gestão do Programa Bolsa Família é descentralizada e compartilhada entre União, estados, Distrito Federal e municípios. Os entes federados trabalham em conjunto para aperfeiçoar, ampliar e fiscalizar a execução. O programa é destinado a famílias em situações de extrema pobreza e pobreza(3. Desde 2004, o PBF encontra-se vinculado ao Ministério do Desenvolvimento Social e Combate à Fome (MDS, mais especificamente à Secretaria Nacional de Renda de Cidadania (Senarc. A inserção das famílias no programa é feita mediante inscrição no Cadastro Único (CadÚnico, de gestão municipal, do qual são selecionadas, de acordo com os critérios do governo federal para o recebimento do benefício(4. Uma das questões mais polêmicas sobre os programas de combate à pobreza é o alcance de sua efetividade. Em pesquisa realizada em 2006(1, em João Pessoa-PB, junto a vinte mães beneficiárias do PBF, os

  13. Impact of the Bolsa Família Program on energy, macronutrient, and micronutrient intakes: Study of the Northeast and Southeast

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    Naiara SPERANDIO

    Full Text Available ABSTRACT Objective: To assess the impact of the Bolsa Família Program on the energy and nutrient intakes of beneficiaries from the Brazilian Northeast and Southeast regions. Methods: The study used data from the 2008-2009 Pesquisa de Orçamento Famíliar, which assessed individual food intake on two nonconsecutive days of individuals aged more than 10 years. Based the personal information booklet, food intake values were transformed into nutritional values (energy and nutrients. Analysis of the impact measure was preceded by propensity score matching, a technique that matches some socioeconomic characteristics of beneficiaries and nonbeneficiaries. Once the score was calculated, the impact of the Bolsa Família Program was estimated by nearest neighbor matching. Results: The program increased energy and macronutrient intakes and decreased calcium and vitamin A, D, E, and C intakes of adolescent beneficiaries in both regions. Adult beneficiaries from the Southeast region increased their fiber, iron, and selenium intakes, and those from the Northeast region decreased their energy, lipid, added sugar, sodium, zinc, vitamin E, and pyridoxine intakes. Conclusion: The results show a positive impact of the program on the energy and macronutrient intakes, and a negative impact on the intakes of most study micronutrients, especially in adolescents, which reinforce the importance of implementing intersectoral actions to improve the nutritional quality of the Bolsa Família Program beneficiaries' diet.

  14. Sepulturas intrusivas Salinar y Chimú en la huaca Herederos Chica, valle de Moche, Perú

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    2003-01-01

    Full Text Available La Huaca Herederos Chica est un site cérémoniel de la Période Initiale et de l’Horizon ancien dans la zone de Caballo Muerto de la vallée de Moche. Les fouilles de Claude Chauchat et de Luis Watanabe en 1970 et 1972 ont amené à la découverte de huit tombes intrusives de la période Salinar et d’une tombe de la période Chimú. Dans les tombes Salinar les corps sont allongés, parfois sur le côté. Les offrandes de céramiques sont disposées à côté, plus rarement sur le corps lui-même. De grands tessons de céramique et des dalles de pierre ont été placés sur les corps, ainsi que sur les vases, comme couvercles. Avec une seule exception, les offrandes de céramique sont des poteries utilitaires portant une décoration minimale et des traces d’utilisation (suie. Dans la tombe Chimú, le corps était replié et probablement attaché à l’intérieur d’un fardeau funéraire. La Huaca Herederos Chica es un sitio ceremonial del Período Inicial y Horizonte Temprano, en el área de Caballo Muerto del valle de Moche. Las excavaciones de Claude Chauchat y Luis Watanabe en 1970 y 1972 han llevado al descubrimiento de ocho tumbas intrusivas del período Salinar y una del período Chimú. Los cuerpos en las tumbas Salinar se encuentran extendidos, a veces en el costado. Los ceramios de ofrendas son dispuestos al lado, o más raramente sobre el cuerpo mismo. Grandes tiestos de cerámica y lajas de piedra fueron colocados sobre el cuerpo y, a modo de tapas, sobre los ceramios. Con una sola excepción, las ofrendas de cerámica son ollas de cocina con decoración mínima y huellas de uso (hollín. En la tumba Chimú, el cuerpo estaba flexionado y probablemente amarrado en un fardo. The Huaca Herederos Chica is a ceremonial Initial period and Early Horizon site within the Caballo Muerto area in the Moche Valley, Peru. Excavations by Claude Chauchat and Luis Watanabe in 1970 and 1972 yielded eight intrusive burials of the Salinar period and

  15. Ressecção de bolsa hiperfuncionante para tratamento de hipotonia ocular crônica: relato de casos Resection of overfiltration bleb for the treatment of chronic ocular hypotony: case reports

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    Sebastião Cronemberger

    2004-08-01

    Full Text Available Relatar os resultados obtidos com a ressecção de bolsa hiperfuncionante pós-trabeculectomia (TREC com mitomicina C (MMC para o tratamento da hipotonia ocular crônica. Cinco pacientes portadores de hipotonia ocular crônica causada por hiperfunção de bolsa fistulante pós- trabeculectomia com mitomicina foram tratados pela ressecção da bolsa. O diagnóstico de hiperfunção da bolsa foi feito com base em critérios estabelecidos pelos autores. A hipotonia ocular foi revertida nos cinco pacientes, sem medicação num seguimento mínimo de cinco e máximo de 26 meses (média de 14,0 ± 7,9 meses. A ressecção da bolsa foi procedimento eficaz para reverter a hipotonia ocular crônica causada pela hiperfunção da mesma pós-trabeculectomia com mitomicina.To present the results of bleb resection for the treatment of overfiltering bleb after trabeculectomy with mitomycin C (MMC associated with chronic ocular hypotony. Five patients with chronic ocular hypotony caused by overfiltering bleb underwent bleb resection. The authors established the criteria for the diagnosis of overfiltering bleb. Ocular hypotony was reversed in all patients without medication. The mean follow-up was 14.0 ± 7.9 months. Bleb resection is a good approach for the treatment of chronic ocular hypotony secondary to overfiltering bleb.

  16. Programa Bolsa Família: uma análise do programa de transferência de renda brasileiro Bolsa Família (Family Grant Programme: an analysis of Brazilian income transfer programme Le programme Bolsa Família (Bourse familiale : analyse du programme brésilien de transfert conditionnel de revenus El programa Bolsa Família: un análisis del programa brasileño de transferencia de ingresos

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    Luciana Mourão

    2012-02-01

    Full Text Available Income transfer programmes are common in various countries and play an important role in combating poverty. This article presents a review of the results of the Bolsa Família (Family Grant Programme, implemented in Brazil by the government of Lula da Silva in 2004. Over the last seven years many evaluations of the programme have been conducted, allowing an overview of its results and its strong and weak points to be mapped. Five central aspects relating to the programme are discussed in article five: (1 programme access, (2 hunger fighting results, (3 programme financial impacts, (4 conditioning factors of education and health, (5 supplementary programs and social mobility. The results of scientific research were presented for each of these aspects, and any of these believed to be convergent or divergent were discussed. As a general result it was concluded that the programme has generated significant results for the country, but there are still some issues that need to be reviewed, such as conditioning factors and the integrated management of the programme.Les programmes de transfert de revenus sont courants dans plusieurs pays et jouent un rôle important dans la lutte contre la pauvreté. Cet article présente un examen des résultats du programme Bolsa Família (Bourse familiale entrepris au Brésil par le gouvernement de Lula da Silva en 2004. Au cours des sept dernières années, de nombreuses évaluations du programme ont été réalisées, ce qui permet d'avoir un aperçu de ses résultats et une vue d'ensemble de ses points forts et de ses points faibles. Cinq aspects clés de ce programme sont abordés dans cet article : (1 l'accès au programme, (2 les résultats en matière de lutte contre la faim, (3 les répercussions financières du programme, (4 les facteurs conditionnels de l'éducation et de la santé, (5 les programmes complémentaires et la mobilité sociale. Des résultats issus de la recherche scientifique ont été pr

  17. Diseño, fabricación y comercialización de bolsas biodegradables

    OpenAIRE

    Díaz Cajiao, Samuel Fernando; Hurtatiz Hernández, Alvaro Roosvel

    2012-01-01

    El plástico y sus derivados, son productos de inmensa utilidad, esto se puede evidenciar en sus aplicaciones en la medicina, la tecnología y en la conveniencia que ofrecen en muchas actividades cotidianas. El problema radica en el uso que se le da, así como la forma como se desecha después de su uso, tal es el caso de las bolsas plásticas usadas en los supermercados. Desarrollar un plan de negocio con miras a determinar y evaluar la viabilidad de crear una empresa cuyo producto estrella s...

  18. Reactor BR2

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2000-07-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

  19. Reactor BR2

    International Nuclear Information System (INIS)

    Gubel, P.

    2000-01-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

  20. Evaluación de tres tipos de empaque (bolsas de polietileno para almacenamiento de guayaba manzana (Psidium guajava var., Klom sali

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    Luis Román Ardila Núñez

    1999-05-01

    Full Text Available La Universidad Nacional de Colombia, a través del Departamento de Ingeniería Agrícola de Santafé de Bogotá, ha venido adelantando investigación sobre manejo postcosecha de productos hortofrutícolas, con miras a minimizar las pérdidas de estos productos y a conservar su calidad. En el presente artículo se muestran los resultados obtenidos de comportamiento del fruto guayaba manzana (Psidium guajava varoKlom Sali, al ser almacenado en frío con bolsas de polietileno de baja densidad de tres tipos: abierto, perforado y cerrado, a una temperatura de 10ºC y humedad relativa de 95 %. Se compararon los resultados durante los días del almacenamiento, tomando como base los índices de madurez del fruto, tales como la pérdida de peso, la intensidad respiratoria, la firmeza, el contenido de ácidos, el contenido de sólidos solubles y el pH. Además, se tomaron datos del almacenamiento de este fruto en bolsas abiertas del mismo tipo, en condiciones ambiente (temperatura 20,1 ºC y humedad relativa de 50,3 %, lo cual se utilizó como testigo. De esta investigación se concluyó que la mejor condición de almacenamiento es en frío con bolsa cerrada, pues el producto conserva mejor su calidad que en los otros dos tipos de empaques evaluados.

  1. Verdades en paralelo bajo la censura: una exploración del híbrido docu-ficción ‘chicas de club’ (1970, de jordi grau

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    Miren GABANTXO URIAGEREA

    2014-07-01

    Full Text Available En este artículo planteamos los modos en que Jordi Grau sortea la censura para crear una película innovadora que desafía tanto las nociones establecidas de cómo hacer cine, como las expectativas de la audiencia. En ‘Chicas de club’ el director se atreve a reflexionar sobre lo que empuja a algunas mujeres a prostituirse en los eufemísticamente denominados “clubs de alterne”. Corren los años setenta y en España, bajo el régimen autoritario del dictador Franco y por lo tanto bajo la censura cinematográfica, el tema de la prostitución está prohibido como tal en el cine. Jordi Grau, en colaboración con el guionista Mario Camus, pone en marcha un arriesgado juego narrativo y visual donde un periodista es el hilo conductor de entrevistas realizadas a gente de la calle y a chicas de alterne, que participan como actrices de su propia vida ficcionada. La película se convierte en un híbrido entre ficción y realidad que seduce al espectador

  2. FACTIBILIDAD DE ALMACENAMIENTO DE SEMILLAS DE AJONJOLÍ (Sesamum indicum L. EN BOLSAS SILOBAG

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    YESID ALEJANDRO MARRUGO-LIGARDO

    Full Text Available El objetivo de esta investigación fue empacar semillas de ajonjolí (Sesamum indicum en bolsas de silobag, evaluando sus características bromatológicas iníciales y después de los treinta y sesenta días de almacenadas a condiciones ambientales y en bodega a 30°C. Las pruebas se hicieron por triplicado, siguiendo los métodos oficiales de análisis; se reportaron los valores promedios. El análisis estadístico indicó que no hubo diferencias significativas respecto a los valores iníciales y los evaluados después de treinta días de almacenado en condiciones ambientales, en cuanto al contenido de fibra (3,98 ± 0,06 vs 4,16 ± 0,13, proteínas (18,86 ± 0,07 vs 19,71 ± 0,89, humedad (5,96 ± 0,06 vs 6,11 ± 0,11, grasa (38,58 ± 0,58 vs 37,49 ± 0,27 y carbohidratos (31,6 ± 0,14 vs 30,76± 0,68. Si se observó algunas variaciones a medida que avanzó el tiempo de la prueba. Se concluyó que las bolsas silobag, se pueden recomendar para empacar ajonjolí y almacenarlo en bodega o dejarlo a la intemperie, dado que protegen al producto contra agentes externos, conservando sus características básicas iníciales, lo cual representa una solución con posibles beneficios económicos para la conservación de este alimento.

  3. Programa Bolsa Família: impacto das transferências sobre os gastos com alimentos em famílias rurais

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    Gisléia Benini Duarte

    2009-12-01

    Full Text Available Programas de transferência condicionada de renda são políticas sociais correntemente empregadas para combater e reduzir a pobreza em diversos países. No curto prazo, esses programas visam aliviar os problemas decorrentes da situação de pobreza, sendo que, no longo prazo, o objetivo é investir no capital humano, quebrando o ciclo intergeracional da pobreza. Estudos têm sido realizados para avaliar os impactos desses programas sobre variáveis como freqüência escolar, trabalho infantil, gastos com alimentação, entre outros. Este trabalho avalia o impacto da transferência de renda do Programa Bolsa Família sobre os gastos com alimentos de famílias rurais. As estimações foram feitas com base no método de Propensity Score Matching (PSM, que corrige para o viés de seleção amostral. Os resultados mostram que o valor médio das despesas anuais para as famílias beneficiárias supera em R$ 246 os gastos totais das famílias não-participantes. Considerando que a média anual recebida por essas famílias é de R$ 278, pode-se inferir que 88% desse valor é utilizado para consumo de alimento. Portanto, o programa de transferência condicionada Bolsa Família exerce um impacto positivo sobre o consumo de alimentos dessas famílias selecionadas.Conditional income transfer programs are social policies currently adopted to reduce poverty in several countries. These conditional transfer schemes have a goal to alleviate some of the consequences of poverty in the short run and increase human capital in the long run changing the intergenerational poverty cycle. Several papers evaluate the impact of income transfer on school attendance, child work and food expenses, among others. This paper analyzes the impact of the Bolsa Família Program on food expenses of rural families. The Propensity Score Method was used to correct sample selection bias. Results show that annual food expenses increased 246 reais in relation to non participant families

  4. Desempenho de estimadores de volatilidade na bolsa de valores de São Paulo

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    Bernardo de Sá Mota

    2004-09-01

    Full Text Available O objetivo deste artigo é avaliar o desempenho de diferentes métodos de extração da volatilidade do Índice da Bolsa de Valores de São Paulo (IBOVESPA tendo como referência a volatilidade realizada. Comparamos modelos da família GARCH com estimadores alternativos baseados em cotações de abertura, fechamento, máximo e mínimo. Os resultados indicam que os estimadores alternativos são tão precisos quanto os modelos do tipo GARCH, apesar de serem muito mais econômicos em termos computacionais.

  5. Aplicación web para la gestión de una bolsa de horas

    OpenAIRE

    Ávila Hernández, Alberto de

    2012-01-01

    El presente documento describe las distintas fases asociadas al diseño y desarrollo de una herramienta que tiene por objetivo informar al usuario del estado de la bolsa de horas de soporte que tiene contratada una determinada empresa. La funcionalidad principal de la aplicación es permitir al usuario conocer cuántas horas de soporte le restan, cuántas han sido consumidas y en qué tareas han sido empleadas, así como obtener distintos tipos de estadísticas con los datos asociados a esta informa...

  6. Verdades en paralelo bajo la censura: una exploración del híbrido docu-ficción Chicas de Club (1970, de Jordi Grau

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    Fernández Guerra, Vanesa

    2010-01-01

    Full Text Available En este artículo planteamos los modos en que Jordi Grau sortea la censura para crear una película innovadora que desafía tanto las nociones establecidas de cómo hacer cine, como las expectativas de la audiencia. En Chicas de club el director se atreve a reflexionar sobre lo que empuja a algunas mujeres a prostituirse en los eufemísticamente denominados “clubs de alterne”. Corren los años setenta y en España, bajo el régimen autoritario del dictador Franco y por lo tanto bajo la censura cinematográfica, el tema de la prostitución está prohibido como tal en el cine. Jordi Grau, en colaboración con el guionista Mario Camus, pone en marcha un arriesgado juego narrativo y visual donde un periodista es el hilo conductor de entrevistas realizadas a gente de la calle y a chicas de alterne, que participan como actrices de su propia vida ficcionada. La película se convierte en un híbrido entre ficción y realidad que seduce al espectador.Abstract in English: In the following paper we look into the ways in which Jordi Grau (1930 navigated the censors to make an innovative film that challenged the established notions of film-making and the audience’s expectations alike. In Chicas de club the director daringly reflected on what pushes some women into prostitution in the euphemistically named "clubs de alterne" (meeting clubs. It is the 1970s and in Spain, under Franco's authoritarian dictatorship - and in a situation of cinematographic censorship - the issue of prostitution is forbidden as a subject for cinema. Jordi Grau, in collaboration with the scriptwriter Mario Camus, sets a risky narrative and visual game in motion in which a journalist is the connecting thread between the interviews carried out with people in the street and the girls in the club, who participate as actresses in the fictional retelling of their very own lives. Thus the film becomes a hybrid of reality and fiction that captivates the viewer.

  7. Programa Bolsa Família e estado nutricional infantil: desafios estratégicos Bolsa Família Program and child nutritional status: strategic challenges

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    Fabiana de Cássia Carvalho Oliveira

    2011-07-01

    Full Text Available Anemia e desnutrição, principais carências nutricionais na infância, têm como principais determinantes os socioeconômicos. Assim, por se tratar da principal política de combate à pobreza, espera-se que o Programa Bolsa Família (PBF promova impacto no estado nutricional infantil. Objetivou-se analisar as diferenças na situação nutricional de crianças cadastradas no PBF de um município da Zona da Mata Mineira. Foram avaliadas 446 crianças com idade entre 6 e 84 meses, sendo que 262 eram beneficiárias e 184 não-beneficiárias. A avaliação nutricional constituiu-se da análise dos parâmetros peso e estatura, através dos índices peso/idade, peso/estatura, estatura/idade e Índice de Massa Corporal/idade, e dos níveis de hemoglobina, com uso do Hemocue. As prevalências de anemia, déficit estatural e obesidade foram 22,6, 6,3 e 5,2%, respectivamente, sendo que não houve diferença estatística entre os beneficiários e não-beneficiários. Inicialmente, o grupo beneficiário apresentava piores condições socioeconômicas, porém, com o recebimento do benefício, os grupos se igualaram financeiramente. É possível que a similaridade dos dois grupos também quanto ao estado nutricional possa ser atribuída ao recebimento do benefício, tanto devido ao incremento financeiro, quanto ao acompanhamento nutricional exigido como condicionalidade do programa.The main nutritional deficiencies during childhood, namely anemia and malnutrition, are predominantly related to socio-economic factors. Thus, as the Bolsa Família Program (BFP is the main policy to combat poverty, it is expected that it will have an impact on child nutrition. The aim was to analyze the differences in the nutritional situation of children registered with the BFP of a municipality located in Zona da Mata of Minas Gerais state. 446 children aged between 6 and 84 months were evaluated, of which 262 were non-beneficiaries and 184 were beneficiaries. Nutritional

  8. Bolsa Família e assimetrias de gênero: reforço ou mitigação?

    Directory of Open Access Journals (Sweden)

    Luana Passos

    Full Text Available Resumo Este artigo tem por objetivo investigar se o programa Bolsa Família contribui para o processo de individualização das mulheres pobres. Para tanto, foi utilizada a técnica de pareamento por escore de propensão, a fim de identificar mulheres e homens não atendidos pelo programa comparáveis a mulheres e homens atendidos. Com base na Pesquisa Nacional por Amostra de Domicílio de 2006, estimaram-se a jornada de trabalho doméstico, a participação no mercado de trabalho e as horas de trabalho remunerado de homens e mulheres. Os resultados não foram conclusivos para participação no mercado de trabalho. Para a jornada de trabalho remunerado, há indícios de que o Programa Bolsa Família reduza as horas trabalhadas de homens e mulheres. Para a jornada de trabalho doméstico, há indicativos de aumento de tempo de cuidado doméstico para mulheres e redução para homens. Os resultados da pesquisa sugerem que o programa reforçaria papéis tradicionais de gênero, não contribuindo para a individualização das mulheres pobres.

  9. Empoderamento das mulheres beneficiárias do Programa Bolsa Família na percepção dos agentes dos Centros de Referência de Assistência Social

    Directory of Open Access Journals (Sweden)

    Nathalia Carvalho Moreira

    2012-04-01

    Full Text Available Este trabalho teve como objetivo analisar o empoderamento das mulheres beneficiárias do Programa de Transferência de Renda, conhecido como "Programa Bolsa Família", na percepção dos agentes sociais dos Centros de Referência de Assistência Social (Cras. Para tanto, realizou-se um estudo de caso múltiplo, tendo como sujeitos de pesquisa 11 gestores de diferentes Cras do estado de Minas Gerais. A partir da técnica de análise de conteúdo, as respostas das perguntas, que compuseram as entrevistas, foram agrupadas de acordo com as categorias Bolsa Família, Cras e Mulher. Os resultados apontam a importância do Cras na execução do Programa Bolsa Família e no processo de empoderamento, pois a convivência e a participação neste local têm contribuído para a conscientização sobre direitos, para a inserção social e para a melhoria do bem-estar das mulheres, fatores evidenciados por intermédio do interesse das mulheres por cursos, oficinas, informações sobre programas sociais e atendimento psicológico. Na percepção dos agentes, foi possível observar melhoria nas condições de vida, nas relações familiares, conscientização e autoestima, implicando reflexos sobre o empoderamento feminino. Portanto, embora sendo um processo lento e embrionário, pode-se dizer que o ciclo do empoderamento das mulheres beneficiárias do Bolsa Família pode ser completado, pois consegue atingir as três dimensões (individual, familiar e comunitária.

  10. Percepções sobre o Programa Bolsa Família na sociedade brasileira

    Directory of Open Access Journals (Sweden)

    Henrique Carlos de Oliveira de Castro

    2009-11-01

    Full Text Available O artigo trata de percepções da sociedade brasileira sobre o Programa Bolsa Família (PBF a partir de uma pesquisa realizada em amostra da população. A pesquisa indicou que a população reconhece o Programa e entende que ele está sendo utilizado de forma adequada, mesmo considerando problemas em sua execução. Houve uma importante diferença entre a opinião daqueles que conhecem beneficiários em relação àqueles que não conhecem, sendo que os primeiros se manifestaram de forma mais positiva em relação aos resultados e mais cautelosos em relação às críticas, conclui que o PBF adquiriu legitimidade junto à sociedade brasileira dado o nível de conhecimento da política e mesmo de apoio à sua existência e argumenta sobre a importância de buscar e considerar a opinião da sociedade como importante elemento de avaliação de políticas públicas.The paper is about perceptions of Brazilian society concerning the cash transfer program Bolsa Família of Brazilian government obtained in a national survey. It indicated that population recognizes the program and understands that it is being used in an appropriated way, even though considering problems in its execution. Important differences occurred in the opinion of those who knew beneficiaries comparing with those who didn't. The first group manifested positive opinion and criticized less then the second. The paper concludes that the program acquired legitimacy in the Brazilian society, considering the knowledge of this policy and supporting its existence. It argues for the importance of searching and considering the public opinion as a fundamental element of public policy assessing.

  11. Efeito do Programa Bolsa Família sobre a oferta de trabalho das mães The impact of the Bolsa Família Program on the labor supply of working mothers

    Directory of Open Access Journals (Sweden)

    Priscilla Albuquerque Tavares

    2010-12-01

    Full Text Available Este artigo investiga a existência de um possível incentivo adverso à oferta de trabalho (participação no mercado e jornada das mães beneficiadas pelo Programa Bolsa Família. Utiliza-se o procedimento de propensity score matching para encontrar mães não atendidas pelo programa comparáveis às mães atendidas, a partir de três grupos de controle. Os resultados apontam a existência de um efeito-renda associado ao valor do benefício, uma vez que quanto maior a transferência recebida, menor o engajamento da mãe no mercado de trabalho. Entretanto, o efeito líquido de ser beneficiário do programa é positivo, indicando a existência de um efeito-substituição, provavelmente decorrente da redução da oferta de trabalho dos filhos, da maior disponibilidade de tempo das mães para trabalhar ou mesmo do estigma em participar do programa.This paper investigates the existence of a possible adverse incentive on the labor supply and weekly working hours of beneficiary mothers of the Bolsa Família Program. Three control groups are analyzed using propensity-score matching to compare non-beneficiary mothers to beneficiary mothers. The results show that there is a wealth effect related to the value of the benefits, given that the larger the benefit, the less active beneficiary mothers are in the labor market. Nonetheless, the net effect is positive, showing that there is a substitution effect due to a reduction in the children´s labor supply, a rise in the mother's available time, as well as decreasing the stigma.

  12. Bolsas de plástico y lazos sociales. Notas de campo sobre reciclaje

    Directory of Open Access Journals (Sweden)

    Cecilia Montero Mórtola

    2011-01-01

    Full Text Available La protección del medio ambiente no sólo consiste en grandes campañas mediáticas y políticas. Existe una ciudadanía silenciosa que ha empezado a modificar actitudes, rescatar viejas costumbres y adaptarlas a distintos espacios de este mundo globalizado (domésticos, informales.... Un cambio cultural donde la reutilización de objetos desechados, a través de actividades artesanales y educativas, sirve para poner en marcha una serie de vínculos y lazos sociales. Estudiando el reciclaje de bolsas de plástico, la antropología puede restituir esos curiosos procesos organizativos, prácticas sociales sólo visibles a partir de un trabajo de campo continuado.

  13. Programa Bolsa Família: uma nova modalidade de biopolítica Family Allowance Program: a new type of biopolitics

    Directory of Open Access Journals (Sweden)

    Rémi Fernand Lavergne

    2012-06-01

    Full Text Available O propósito deste artigo é mostrar como o Programa Bolsa Família (PBF remete a uma forma de biopolítica nos termos evocados por Michel Foucault e inscreve-se numa perspectiva de normalização, funcionando pela norma e pela regulamentação. Busca também evidenciar como o Serviço Social e a educação "por toda a vida" têm um importante papel nos processos de subjetivação e de produção de subjetividades com vistas a incidir sobre a conduta das populações indigentes e marginalizadas.The purpose of this article is to show how the Family Allowance Program (Programa Bolsa Família - PBF refers to a form of biopolitics such as evoked by Foucault. Besides illustrating how the PBF is inscribed in a normalizing perspective, the article seeks to show how the Social Services and the Education "for a lifetime" play an important role in terms of building up subjectivity processes and subjective opinions and feelings whose aim is to guide and control the poor and marginalized populations' behavior.

  14. PARR-2: reactor description and experiments

    International Nuclear Information System (INIS)

    Wyne, M.F.; Meghji, J.H.

    1990-12-01

    PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

  15. Reactor BR2: Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2000-01-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. A safety audit was conduced by the IAEA, the conclusions of which demonstrated the excellent performance of the plant in terms of operational safety. In 1999, the CALLISTO facility was extensively used for various programmes involving LWR pressure vessel materials, IASCC of LWR structural materials, fusion reactor materials and martensic steels for use in ADS systems. In 1999, BR2's commercial programmes were further developed

  16. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

    2012-03-15

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  17. The elusive character of discontinuous deep-water channels: New insights from Lucia Chica channel system, offshore California

    Science.gov (United States)

    Maier, K.L.; Fildani, A.; Paull, C.K.; Graham, S.A.; McHargue, T.R.; Caress, D.W.; McGann, M.

    2011-01-01

    New high-resolution autonomous underwater vehicle (AUV) seafloor images, with 1 m lateral resolution and 0.3 m vertical resolution, reveal unexpected seafloor rugosity and low-relief (thalwegs were interpreted originally from lower-resolution images, but newly acquired AUV data indicate that a single sinuous channel fed a series of discontinuous lower-relief channels. These discontinuous channels were created by at least four avulsion events. Channel relief, defined as the height from the thalweg to the levee crest, controls avulsions and overall stratigraphic architecture of the depositional area. Flowstripped turbidity currents separated into and reactivated multiple channels to create a distributary pattern and developed discontinuous trains of cyclic scours and megaflutes, which may be erosional precursors to continuous channels. The diverse features now imaged in the Lucia Chica channel system (offshore California) are likely common in modern and ancient systems with similar overall morphologies, but have not been previously mapped with lower-resolution detection methods in any of these systems. ?? 2011 Geological Society of America.

  18. EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

    International Nuclear Information System (INIS)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  19. [The impact of conditional cash transfers on health status: the Brazilian Bolsa Familia Programme].

    Science.gov (United States)

    Rivera Castiñeira, Berta; Currais Nunes, Luis; Rungo, Paolo

    2009-01-01

    Conditional cash transfers are becoming the standard approach to reducing poverty levels; the Brazilian Bolsa Familia Program, in particular, is the largest program of this kind, and the evaluation of its impact allows for drawing some interesting conclusions, which may apply to other countries. In this paper, the lack of positive results in terms of both health status and modification of unhealthy habits is underlined. Among different causes, which are discussed here, the existence of barriers on the supply side appears as the most important limitation for obtaining better results. The positive impact of this program on both education and poverty reduction however, allows for predicting improvements in health status in the long run.

  20. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2001-01-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

  1. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2001-04-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  2. Safety features of TR-2 reactor

    International Nuclear Information System (INIS)

    Tuerker, T.

    2001-01-01

    TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

  3. Bolsa Família (Family Grant Programme: an analysis of Brazilian income transfer programme

    Directory of Open Access Journals (Sweden)

    Luciana Mourão

    2012-06-01

    Full Text Available Income transfer programmes are common in various countries and play an important role in combating poverty. This article presents a review of the results of the Bolsa Família (Family Grant Programme, implemented in Brazil by the government of Lula da Silva in 2004. Over the last seven years many evaluations of the programme have been conducted, allowing an overview of its results and its strong and weak points to be mapped. Five central aspects relating to the programme are discussed in article five: (1 programme access, (2 hunger fighting results, (3 programme financial impacts, (4 conditioning factors of education and health, (5 supplementary programs and social mobility. The results of scientific research were presented for each of these aspects, and any of these believed to be convergent or divergent were discussed. As a general result it was concluded that the programme has generated significant results for the country, but there are still some issues that need to be reviewed, such as conditioning factors and the integrated management of the programme.

  4. [Intersectoral, convergent and sustainable actions: the challenges of the "Bolsa Família" program in Manguinhos shantytown in Rio de Janeiro].

    Science.gov (United States)

    Magalhães, Rosana; Coelho, Angela Virginia; Nogueira, Milena Ferreira; Bocca, Cláudia

    2011-11-01

    Some studies have revealed the impact of the family welfare allowance based on the fulfillment of certain conditions on improving living conditions and access to health and education services in different countries. However, gaps persist relating to the evaluation of the benefits of such programs among the groups that have greater difficulty in gaining access to public services or advances in the quality of education and school performance. Moreover, there is limited evidence of adequacy of the program to the respective contexts of implementation, levels of adhesion and local cooperation and strategies adopted for integration with other social policy programs. The scope of this article is to discuss the findings of the study of the implementation of the "Bolsa Familia" in the Manguinhos shantytown area in Rio de Janeiro conducted in 2007 and 2008 based on semi-structured interviews with program officials and local stakeholders. In conclusion, the study shows that the sustainability of "Bolsa Familia" actions to reduce poverty and promote health equity calls for strengthening the vertical and horizontal communication channels between government levels, public managers and civil associations, recognition of the complexity of the local social demands and an intersectoral agenda.

  5. A bolsa na mediação "estar ostomizado" - "estar profissional": análise de uma estratégia pedagógica La bolsa como mediadora entre "estar ostomizado" y "estar profesional": análisis de una estratégia pedagógica The pouch mediating the relation between "being an ostomized person" and "being professional": analysis of a pedagogic strategy

    Directory of Open Access Journals (Sweden)

    Vera Lúcia Conceição de Gouveia Santos

    2000-07-01

    Full Text Available Este estudo analisou a (reconstrução das significações sobre a ostomia, o ostomizado, o cuidar em enfermagem e o papel profissional de 30 enfermeiros que utilizaram bolsa coletora, em experiência pedagógica, durante os Cursos de Estomaterapia. A análise dos depoimentos revelou dois grandes eixos discursivos: "estar ostomizado" e "estar profissional". O enfermeiro, tendo por mediação o uso da bolsa coletora, vivencia o "estar ostomizado" por meio de violações da identidade e qualidade de vida, perpassadas por transformações desde papéis às relações com o outro. A mobilização de conteúdos simbólicos e afetivos acerca do "estar ostomizado" gera uma crise de significação do "estar profissional", até então caracterizado por um cuidar fragmentado. Re-conhecendo o cuidar passado como um fazer técnico voltado principalmente para a ostomia-bolsa, o aluno-enfermeiro projeta um cuidar futuro mais holístico do ser humano portador de uma ostomia, incorporando as dimensões afetivas, simbólicas e relacionais.El estudio es acerca de la (reconstrucción de los significados sobre el estoma, el ostomizado, el cuidado de enfermería y el rol profesional, hecha por 30 enfermeros que usaran bolsas de drenaje, como una estrategia pedagógica en los Cursos de Terapia Enterostomal. El análisis de los discursos reveló dos grandes categorias: "estar ostomizado" y "estar profesional". El enfermero, teniendo por mediación la bolsa, vive el "estar ostomizado" con sentimientos de transgresión de la identidad, de la calidad de vida, cambios de roles y de sus relaciones sociales. La movilización de los contenidos simbólicos y afectivos acerca del "estar ostomizado" resulta en una crisis de significado del "estar profesional" caracterizado por el cuidar fragmentado. Reconociendo el cuidar pasado como un hacer técnico dirigido principalmente para el estoma-bolsa, el enfermero proyecta un cuidar futuro del ostomizado más humanizado, agregando

  6. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2002-01-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system

  7. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2002-04-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

  8. Inventory and Evaluation of Cultural Resources, Bolsa Chica Mesa and Huntington Beach Mesa, Orange County, California

    Science.gov (United States)

    1989-09-30

    Excelentisimo Conde de Monterey, Virrey Que Era dela Nueva Espana. In Monarchia Indiana, edited by J. de Torquemada, pp. 693-725. Madrid. 101 102 Baumhoff, M...biological bacterias , this includes the destruction of canyons, hills, mountains and the flora and fauna in these areas. Road construction, real

  9. Impacto do Programa Bolsa Família sobre a frequência escolar: o caso da agricultura familiar no Nordeste do Brasil

    Directory of Open Access Journals (Sweden)

    Raul da Mota Silveira Melo

    2010-09-01

    Full Text Available O objetivo deste trabalho é avaliar o impacto do programa de transferência de renda condicionada Bolsa Família sobre a frequência escolar de crianças e adolescentes de cinco a 14 anos na agricultura familiar dos estados de Pernambuco, Ceará, Sergipe e Paraíba. Nessa investigação, o trabalho faz uso de dados primários (pesquisa de campo e dados secundários (PNAD, 2005 para obter estimativas de propensity score. Os resultados indicam que, de forma geral, o programa eleva a frequência escolar das referidas crianças no intervalo de 5,4 a 5,9 pontos percentuais. Contudo, há importantes diferenças quando se considera meninas e meninos separadamente, sendo o programa eficaz no primeiro caso e ineficaz no segundo. Ou seja, apesar da avaliação positiva para as meninas, não parece haver efeito do programa sobre a frequência escolar dos meninos, o que pode estar associado a diferenças de gênero nos custos de oportunidades do investimento em capital humano no meio rural.The main proposal of this study was to evaluate the impact of the Bolsa Família conditioned public cash transfer program on the school presence among the children and adolescents from five to 14 years, in the Brazilian states of Pernambuco, Ceará, Sergipe and Paraíba. The work uses both primary and secondary data (PNAD, 2005 to build two different control groups used for propensity score estimative matching with children from families that received income from the federal program. For all studied groups the impact of the Bolsa Família was positive, in other words, the results indicate that the program increases the school presence by 5,6 points. But the results still suggest there is difference between gender, with the program being effective for girls, but not for boys. This probably is related to the gender difference in the opportunity cost of human capital investment in Brazil rural northeast.

  10. BR2 Reactor: Introduction

    International Nuclear Information System (INIS)

    Moons, F.

    2007-01-01

    The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3 He, 6 Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

  11. Inicio de relaciones sexuales con penetración y factores asociados en chicos y chicas de México de 14-19 años de edad con escolarización en centros públicos

    Directory of Open Access Journals (Sweden)

    Leonor Rivera-Rivera

    2016-01-01

    Conclusiones: En México, el IRSP se presenta a edad más temprana en los chicos. Además, los hallazgos del presente estudio demuestran que la edad de IRSP y los factores asociados son diferentes en los chicos y las chicas. Las creencias de género y socioculturales influyen de manera importante en la edad de IRSP.

  12. Efeitos do Programa Bolsa Família na fecundidade das beneficiárias

    Directory of Open Access Journals (Sweden)

    Patrícia Simões

    2012-12-01

    Full Text Available Procuramos verificar se o Programa Bolsa Família contribui para aumentar a fecundidade entre as beneficiárias, visto que o aumento no tamanho da família, até certo limite, leva ao aumento dos benefícios. Utilizamos um modelo de contagem no qual testamos e tratamos a possibilidade de endogeneidade da variável de política por dois métodos distintos (dois-estágios estilo Heckman e GMM, além de incluir diversos cofatores da PNDS (2006. Os resultados mostram que o PBF não apresentou este efeito, pelo menos no início do programa. Pelo contrário, beneficiárias pareciam mais inclinadas a trocar quantidade por qualidade do que não beneficiárias elegíveis ao programa.

  13. A família nas políticas sociais: o caso do Programa Bolsa Família

    OpenAIRE

    Marcelo Couto Dias

    2013-01-01

    Nos países ocidentais, os últimos anos têm sido marcados por uma crescente redescoberta do valor da família e das microssolidariedades. Prova disso é o aparecimento da família tanto nas discussões das políticas sociais, quanto nos processos de formulação das mesmas. No contexto brasileiro recente, ganhou destaque a criação e expansão do Bolsa Família, um programa de transferência direta de renda com condicionalidades que, em 2012, tinha entre os seus beneficiários mais de 13 mi...

  14. DESAFIOS PARA A COORDENAÇÃO INTERGOVERNAMENTAL DO PROGRAMA BOLSA FAMÍLIA

    Directory of Open Access Journals (Sweden)

    Claudia Regina Baddini Curralero

    2011-08-01

    Full Text Available The paper examines the intergovernmental coordination of the Bolsa Família Program (PBF, given its goal to tackle poverty in acountry with deep social and regional inequality. It seeks to qualify the debate on the centralization of cash transfer programs in Brazil through the analysis of intergovernmental relations adopted under the three main dimensions of the PBF - cash transfer, monitoring of conditionalities and articulation of complementary programs - considering the federative implications derived from the intersectoral perspective that drives the Program. Two categories of challenges are highlighted. The first one relates to the need for greater investment in space and opportunities for intergovernmental negotiation, especially in the dimension of income transfer, which initially was characterized by centralization. In the second, the intergovernmental matter demands the organization of a coordinated national strategy for the articulation of complementary programs, suggesting greater involvement of states in the regional coordination of the Program

  15. Application of MCNPX 2.7.D for reactor core management at the research reactor BR2

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, Edgar

    2011-01-01

    The paper discusses application of the Monte Carlo burn up code MCNPX 2.7.D for whole core criticality and depletion analysis of the Material Testing Research Reactor BR2 at SCK-CEN in Mol, Belgium. Two different approaches in the use of MCNPX 2.7.D are presented. The first methodology couples the evolution of fuel depletion, evaluated by MCNPX 2.7.D in an infinite lattice with a steady-state 3-D power distribution in the full core model. The second method represents fully automatic whole core depletion and criticality calculations in the detailed 3-D heterogeneous geometry model of the BR2 reactor. The accuracy of the method and computational time as function of the number of used unique burn up materials in the model are being studied. The depletion capabilities of MCNPX 2.7.D are compared vs. the developed at the BR2 reactor department MCNPX & ORIGEN-S combined method. Testing of MCNPX 2.7.D on the criticality measurements at the BR2 reactor is presented. (author)

  16. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    International Nuclear Information System (INIS)

    Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

    2014-01-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

  17. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    Energy Technology Data Exchange (ETDEWEB)

    Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

    2014-07-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

  18. Core monitoring at the WNP-2 reactor

    International Nuclear Information System (INIS)

    Skeen, D.R.; Torres, R.H.; Burke, W.J.; Jenkins, I.; Jones, S.W.

    1992-01-01

    The WNP-2 reactor is a 3,323-MW(thermal) boiling water reactor (BWR) that is operated by the Washington Public Power Supply System. The WNP-2 reactor began commercial operation in 1984 and is currently in its eighth cycle. The core monitoring system used for the first cycle of operation was supplied by the reactor vendor. Cycles 2 through 6 were monitored with the POWERPLEX Core Monitoring Software System (CMSS) using the XTGBWR simulation code. In 1991, the supply system upgraded the core monitoring system by installing the POWERPLEX 2 CMSS prior to the seventh cycle of operation for WNP-2. The POWERPLEX 2 CMSS was developed by Siemens Power Corporation (SPC) and is based on SPC's advanced state-of-the-art reactor simulator code MICROBURN-B. The improvements in the POWERPLEX 2 system are possible as a result of advances in minicomputer hardware

  19. Historia singular de una chica inmigrante a su paso por el primer ciclo de la Educación Secundaria

    Directory of Open Access Journals (Sweden)

    Dolores RODRÍGUEZ MARTÍNEZ

    2012-01-01

    Full Text Available El estudio que se presenta es resultado de una investigación realizada en el ámbito institucional de un centro de Secundaria con la intención de comprender la singularidad de la vida de una chica adolescente inmigrante desde su propia voz y a partir de su experiencia escolar, entendiendo ésta en su dimensión global, en la que se conjugan circunstancias personales, culturales, sociales, institucionales y académicas en constante diálogo, emergiendo una propia e idiosincrásica identidad. Situado en un paradigma naturalista y bajo un diseño etnográfico, el estudio adopta la forma de narrativa temporal para finalizar con unos apuntes para la reflexión que, a modo de conclusiones, evidencian la necesidad de contemplar a los estudiantes más allá de su rendimiento académico y de un patrón escolar homogéneo.

  20. TA-2 Water Boiler Reactor Decommissioning Project

    International Nuclear Information System (INIS)

    Durbin, M.E.; Montoya, G.M.

    1991-06-01

    This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

  1. EBR-2 [Experimental Breeder Reactor-2] test programs

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  2. Mulher e família no Programa Bolsa-Escola: maternidades veiculadas e instituídas pelos anúncios televisivos Woman and family at the Bolsa-Escola Program: maternities propagated and instituted through TV advertisements

    Directory of Open Access Journals (Sweden)

    Carin Klein

    2007-12-01

    Full Text Available Este artigo problematiza alguns processos de produção e veiculação de representações de maternidade, tomando como referência o Programa Nacional Bolsa-Escola, e insere-se no campo da teorização cultural, principalmente na perspectiva dos Estudos Culturais e dos Estudos Feministas, nas vertentes que têm proposto uma aproximação crítica com a análise pós-estruturalista. Para a operacionalização do trabalho, selecionei um conjunto de anúncios televisivos que divulgaram o Programa à população no primeiro ano de sua implantação. Exploro os anúncios com o intuito de analisar os diferentes modos de representar e significar a maternidade. Discuto como se organiza e divulga, no âmbito do Programa, um conjunto de ensinamentos e propostas a serem desenvolvidas, principalmente na família, a fim de buscar (recolocar, sobretudo, as mulheres-mães e a educação das crianças no centro desses debates.This work discusses and questions some processes of production and propagation of maternity representations, having the National Bolsa-Escola Program as its starting point, and localized in the field of cultural theory, mainly from the perspectives of both Cultural Studies and Feminist Studies, with a critical approximation to the post-structuralist analysis. In order to carry out the work, I have selected a series of television advertisements used to publicize the Program in its first year of implementation. I have explored these advertisements in order to analyze the different ways through which maternity has been represented and meant. I have discussed how a set of teachings and proposals was publicized in the Program so as to be mainly developed by the families, thus relocating women/mothers and children’s education into the center of those debates.

  3. Evolução do Programa Bolsa Família: Brasil e estados do Nordeste 2004-2009

    OpenAIRE

    Queiroz, Silvana Nunes de; Remy, Maria Alice Pestana de Aguiar; Pereira, Júlia Modesto Pinheiro Dias; Silva, Luis Abel da

    2010-01-01

    Este artigo analisa a evolução no número de beneficiários e no valor do repasse do Programa Bolsa Família (PBF). Para tanto, são feitas considerações sobre o conceito de pobreza e as principais alterações na concepção do PBF. O estudo tem como recorte temporal os anos de 2004 a 2009, e recorte espacial o Nordeste brasileiro, região com os piores indicadores sociais e demográficos do país. A fonte de dados foi a Matriz de Informação Social do MDS (Ministério do Desenvolvimento Social), que apo...

  4. Intersetorialidade, convergência e sustentabilidade: desafios do programa Bolsa Família em Manguinhos, RJ Intersectoral, convergent and sustainable actions: the challenges of the "Bolsa Família" program in Manguinhos shantytown in Rio de Janeiro

    Directory of Open Access Journals (Sweden)

    Rosana Magalhães

    2011-11-01

    Full Text Available Alguns estudos têm revelado o impacto de programas de transferência condicionada de renda na melhoria das condições de vida e no acesso a serviços básicos de saúde e educação em diferentes países. No entanto, persistem lacunas no que se refere à avaliação dos benefícios de tais intervenções entre os grupos que apresentam maiores dificuldades em acessar serviços públicos ou dos avanços na qualidade do ensino e desempenho escolar. Além disso, existem poucas evidências sobre a adequação das ações aos respectivos contextos de implementação, níveis de adesão e cooperação, local e estratégias adotadas para a integração com as demais políticas de proteção social. O artigo discute os resultados da pesquisa avaliativa sobre a implementação do programa de transferência condicionada de renda Bolsa Família em Manguinhos (RJ realizada entre os anos de 2007 e 2008. Foram realizadas entrevistas com gestores das secretarias municipais de assistência social, saúde e educação e agentes implementadores locais. Em Manguinhos, a sustentabilidade das ações voltadas à redução da pobreza e promoção da saúde envolve o fortalecimento de canais de interlocução entre níveis de governo, gestores públicos e associações civis, reconhecimento da complexidade das demandas sociais locais e pactuação de uma agenda intersetorial.Some studies have revealed the impact of the family welfare allowance based on the fulfillment of certain conditions on improving living conditions and access to health and education services in different countries. However, gaps persist relating to the evaluation of the benefits of such programs among the groups that have greater difficulty in gaining access to public services or advances in the quality of education and school performance. Moreover, there is limited evidence of adequacy of the program to the respective contexts of implementation, levels of adhesion and local cooperation and

  5. BR2 Reactor: Irradiation of fuels

    International Nuclear Information System (INIS)

    Verwimp, A.

    2005-01-01

    Safe, reliable and economical operation of reactor fuels, both UO 2 and MOX types, requires in-pile testing and qualification up to high target burn-up levels. In-pile testing of advanced fuels for improved performance is also mandatory. The objectives of research performed at SCK-CEN are to perform Neutron irradiation of LWR (Light Water Reactor) fuels in the BR2 reactor under relevant operating and monitoring conditions, as specified by the experimenter's requirements and to improve the on-line measurements on the fuel rods themselves

  6. Avaliação de bolsas de produtividade em pesquisa do CNPq e medidas bibliométricas: correlações para todas as grandes áreas

    Directory of Open Access Journals (Sweden)

    Jacques Wainer

    Full Text Available Este trabalho estuda as correlações entre decisões tomadas no fim de 2009, sobre renovação ou não de bolsas de produtividade em pesquisa do CNPq e medidas bibliométricas. Para cada nível da bolsa e para cada subárea, calculamos a correlação da decisão em subir o pesquisador de nível, mantê-lo no nível original ou rebaixá-lo, com várias medidas bibliométricas, como produção total (artigos, conferências, livros e capítulos de livros, produção nos últimos 5 anos, produção indexada no Web of Science, citações recebidas por artigo, citações recebidas por artigo escrito nos últimos 5 anos, índice H, etc. Os dados de citações foram extraídos tanto do Google Scholar como do Web of Science. As correlações de cada subárea são agrupadas em cada uma das 8 grandes áreas do CNPq (Ciências Agrícolas, Ciências Biológicas, Ciências Exatas, Ciências Humanas, Ciências da Saúde, Ciências Sociais, Engenharia e Artes. Indicamos quais são as métricas bibliométricas com maior correlação, com as decisões do CNPq para cada nível e para cada uma das grandes áreas. Discutimos algumas grandes áreas nas quais parece haver uma maior coerência, através dos vários níveis da bolsa entre as métricas mais correlacionadas com as decisões.

  7. Ageing management of the BR2 research reactor

    International Nuclear Information System (INIS)

    Verpoortem, J. R.; Van Dyck, S.

    2014-01-01

    At the Belgian nuclear research centre (SCK.CEN) several test reactors are operated. Among these, Belgian Reactor 2 (BR2) is the largest Material Test Reactor (MTR). This water-cooled, beryllium moderated reactor with a maximum thermal power of 100 MW became operational in 1962. Except for two major refurbishment campaigns of one year each, this reactor has been operated continuously over the past 50 years, with a frequency of 5-12 cycles per year. At present, BR2 is used for different research activities, the production of medical isotopes, the production of n-doped silicon and various training and education activities. (Author)

  8. Estudio de la biodegradación de bolsas oxo - biodegradables en agua dulce y salada, simulando condiciones ambientales de Costa, Sierra y Oriente Ecuatoriano.

    OpenAIRE

    Escobar Silva, Nelly Jacqueline

    2014-01-01

    This work allowed the study of biodegradation of four types of oxo-biodegradable bags by environmental conditions simulated in fresh and salt water of the Costa, Sierra and Oriente regions. El presente trabajo permitió el estudio de la biodegradación de cuatro tipos de bolsas oxo-biodegradables mediante condiciones ambientales simuladas en el agua dulce y salada de las regiones Costa, Sierra y Oriente ecuatoriano.

  9. Una revisión sobre las escuelas de educación secundaria para chicas en Inglaterra en el siglo XIX.

    Directory of Open Access Journals (Sweden)

    Cristina Yanes Cabrera

    2011-08-01

    Full Text Available El contexto general que caracterizó la educación secundaria en sus orígenes  en Europa, ha venido siendo un objeto de estudio bastante extendido en la Historia de la educación. Dentro de este nivel educativo y en el contexto específico de Inglaterra, este trabajo se plantea dar a conocer las principales características de la educación ofrecida a las chicas frente a la de los chicos. Para ello, se ha llevado a cabo a lo largo de todo el siglo diecinueve un estudio de las principales instituciones femeninas destinadas a la educación secundaria, así como de su currículo y de su organización. Se pretende, de esta manera, dejar constancia del carácter y finalidad del progresivo acceso de la mujer inglesa a la educación secundaria.

  10. El diseño de un juego autogenerativo de títulos de bolsa

    Directory of Open Access Journals (Sweden)

    Guillermo Buenaventura Vera

    2004-01-01

    Full Text Available Contando el establecimiento de la función generadora de precios aleatorios, se desarrolla la metodología básica para la construcción de un modelo simulador de juego de bolsa que sea capaz de generar las propias variaciones de los precios de las acciones. El artículo realiza la presentación estructurada de modelo, partiendo de las bases teóricas para la elaboración de la formulación. La generación de números aleatorios distribuidos mediante la función normal estándar se construye a partir de la función uniforme generadora de números aleatorios (RAND. Las consideraciones de programación de computadores, así como una estructura básica de la misma, son tratadas enfocando tanto aplicaciones individuales del juego de simulación, como aplicaciones en red.

  11. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  12. O Programa Bolsa Família e os Pobres "Não Merecedores": poder discricionário e os limites da consolidação de direitos sociais

    NARCIS (Netherlands)

    Eiró de Oliveira, F.H.

    2017-01-01

    Sendo o maior programa de transferência condicionada de renda do mundo (em número absoluto de pessoas assistidas), o impacto do Programa Bolsa Família (PBF) vai além da redução de vulnerabilidades materiais. Em minhas pesquisas, constatei que o programa pode representar o único contato positivo das

  13. BR2 reactor neutron beams

    International Nuclear Information System (INIS)

    Neve de Mevergnies, M.

    1977-01-01

    The use of reactor neutron beams is becoming increasingly more widespread for the study of some properties of condensed matter. It is mainly due to the unique properties of the ''thermal'' neutrons as regards wavelength, energy, magnetic moment and overall favorable ratio of scattering to absorption cross-sections. Besides these fundamental reasons, the impetus for using neutrons is also due to the existence of powerful research reactors (such as BR2) built mainly for nuclear engineering programs, but where a number of intense neutron beams are available at marginal cost. A brief introduction to the production of suitable neutron beams from a reactor is given. (author)

  14. Fission product release from SLOWPOKE-2 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

    1994-12-31

    Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

  15. Focus and coverage of Bolsa Família Program in the Pelotas 2004 birth cohort.

    Science.gov (United States)

    Schmidt, Kelen H; Labrecque, Jeremy; Santos, Iná S; Matijasevich, Alicia; Barros, Fernando C; Barros, Aluisio J D

    2017-03-30

    To describe the focalization and coverage of Bolsa Família Program among the families of children who are part of the 2004 Pelotas birth cohort (2004 cohort). The data used derives from the integration of information from the 2004 cohort and the Cadastro Único para Programas Sociais do Governo Federal (CadÚnico - Register for Social Programs of the Federal Government), in the 2004-2010 period. We estimated the program coverage (percentage of eligible people who receive the benefit) and its focus (proportion of eligible people among the beneficiaries). We used two criteria to define eligibility: the per capita household income reported in the cohort follow-ups and belonging to the 20% poorest families according to the National Economic Indicator (IEN), an asset index. Between 2004 and 2010, the proportion of families in the cohort that received the benefit increased from 11% to 34%. We observed an increase in all wealth quintiles. In 2010, by income and wealth quintiles (IEN), 62%-72% of the families were beneficiaries among the 20% poorest people, 2%-5% among the 20% richest people, and about 30% of families of the intermediate quintile. According to household income (minus the benefit) 29% of families were eligible in 2004 and 16% in 2010. By the same criteria, the coverage of the program increased from 43% in 2004 to 71% in 2010. In the same period, by the wealth criterion (IEN), coverage increased from 29% to 63%. The focalization of the program decreased from 78% in 2004 to 32% in 2010 according to income, and remained constant (37%) according to the IEN. Among the families of the 2004 cohort, there was a significant increase in the program coverage, from its inception until 2010, when it was near 70%. The focus of the program was below 40% in 2010, indicating that more than half of the beneficiaries did not belong to the target population. Descrever a focalização e a cobertura do Programa Bolsa Família nas famílias de crianças que fazem parte da coorte

  16. Staphylococcus spp. in the oral cavity and periodontal pockets of chronic periodontitis patients Staphylococcus spp. na cavidade bucal e na bolsa periodontal de indivíduos com periodontite crônica

    Directory of Open Access Journals (Sweden)

    Jussara Cia S. Loberto

    2004-06-01

    outras infecções pode predispor o aumento do número de Staphylococcus spp. na boca, pois estes adquirem facilmente resistência aos antibióticos, podendo resultar em superinfecção. O objetivo deste estudo foi verificar a presença de Staphylococcus spp. na cavidade bucal e nas bolsas periodontais de pacientes com periodontite crônica; identificar as cepas isoladas; verificar a relação entre a presença de Staphylococcus spp. na cavidade bucal e presença de bolsa periodontal. Participaram deste estudo 88 pacientes, entre 25 e 60 anos de idade e apresentando periodontite crônica, com pelo menos dois sítios com profundidade de sondagem maior ou igual a 5mm. Após anamnese e exame clínico periodontal foram feitas coletas de material da bolsa periodontal com cones de papel e da cavidade bucal por meio de bochechos. Do total de pacientes 37,50% apresentaram Staphylococcus spp. na bolsa periodontal e 61,36% na cavidade bucal, sendo que 27,27% apresentaram a bactéria nos 2 sítios. S. epidermidis foi a espécie mais prevalente para bolsa periodontal (15,9% e cavidade bucal (27,27%. Não houve diferença estatística significante quanto à presença desses microrganismos entre as faixas etárias e aumento da profundidade de sondagem. A presença de bactérias oportunistas na cavidade bucal pode representar dificuldades para a manutenção do tratamento periodontal.

  17. The "local economy" effect of social transfers : an empirical assessment of the impactof th Bolsa Familia program on local productive structure and economic growth

    OpenAIRE

    Rougier, E.; Combarnous, F.; Fauré, Yves-André

    2018-01-01

    Social transfers impact local economic growth through local demand multiplier and local productive structures. Using original data on productive structures, growth determinants and Bolsa Familia conditional transfers (BFP) for the 184 municipalities of the Brazilian state of Ceará during 2003–10, we show that the positive impact of the transfers on local growth is in fact conditional on the direction of local economic structure transformation. Indeed, transfers did spur light manufactur...

  18. El acceso de la mujer a cargos de toma de decisiones en las empresas colombianas que cotizan en bolsa

    Directory of Open Access Journals (Sweden)

    Lina Marrugo-Salas

    2016-01-01

    Full Text Available En este artículo se analiza el fenómeno del techo de cristal y se presenta un panorama de la proporción de mujeres en los más altos cargos directivos de 76 empresas colombianas que cotizan en bolsa. La investigación fue de carácter documental mediante la recolección, procesamiento y análisis de la información pública disponible. Los resultados muestran la baja participación de las mujeres en dichos puestos de responsabilidad, por lo que se propone aumentar el rol que tienen en las empresas mediante el desarrollo de programas de responsabilidad social.

  19. Pressurized water reactor simulator. Workshop material. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

  20. Condicionalidades em saúde do programa Bolsa Família – Brasil: uma análise a partir de profissionais da saúde

    Directory of Open Access Journals (Sweden)

    Alice Teles de Carvalho

    2014-12-01

    Full Text Available Este estudo apresenta a percepção de profissionais de equipes de Saúde da Família de municípios do Nordeste do Brasil acerca das mudanças na vida das famílias participantes do programa Bolsa Família, da relação destas com os serviços de saúde e do impacto na dinâmica de trabalho dos profissionais, a partir do acompanhamento das condicionalidades de saúde do programa Bolsa Família. As informações foram obtidas por meio de entrevistas semiestruturadas e encontros de grupo focal. Os profissionais acreditam que o programa ocasionou mudanças favoráveis na vida das famílias participantes, como a redução da pobreza, o aumento da frequência escolar das crianças e mudanças positivas na relação entre as famílias participantes e os serviços de saúde. No entanto, relataram dificuldades de caráter organizacional no acompanhamento das condicionalidades, sobretudo devido ao aumento da demanda de trabalho. É importante que as condicionalidades de saúde proporcionem oportunidades para a realização de ações que visem ao empoderamento e autonomia dos sujeitos quanto ao autocuidado e desenvolvimento da cidadania.

  1. Advances in Reactor Physics, Mathematics and Computation. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, Volume 2, are divided into 7 sessions bearing on: - session 7: Deterministic transport methods 1 (7 conferences), - session 8: Interpretation and analysis of reactor instrumentation (6 conferences), - session 9: High speed computing applied to reactor operations (5 conferences), - session 10: Diffusion theory and kinetics (7 conferences), - session 11: Fast reactor design, validation and operating experience (8 conferences), - session 12: Deterministic transport methods 2 (7 conferences), - session 13: Application of expert systems to physical aspects of reactor design and operation.

  2. O Impacto das Regras do Programa Bolsa Família Sobre a Fecundidade das Beneficiárias

    Directory of Open Access Journals (Sweden)

    Luis Antonio Winck Cechin

    2015-09-01

    Full Text Available Este trabalho investiga um possível incentivo do Programa Bolsa Família ao aumento da fecundidade de suas beneficiárias em decorrência de suas regras, dado que a quantidade de recursos transferidos depende do número de filhos da família. O diferencial deste estudo reside na análise desse impacto em um maior período de exposição das beneficiárias aos efeitos do PBF. Aplica-se o algoritmo de seleção de covariadas proposto por Imbens (2014 e o método de Propensity Score Matching. Os resultados apontaram que o PBF gera pequeno incentivo à geração do segundo filho, sendo que as regiões Centro-Oeste e Nordeste apresentaram os maiores valores de impacto.

  3. Relación entre la presencia de bolsas periodontales y las alteraciones del perfil lipídico en pacientes con ateroesclerosis

    OpenAIRE

    Ruiz Alvarez, Carmen; Pareja Vásquez, María del Carmen

    2006-01-01

    Objetivo: determinar la relación entre la presencia de bolsas periodontales y la alteración en los valores del perfil lipídico (niveles plasmáticos de colesterol, triglicéridos, HDL, LDL) en pacientes con ateroesclerosis. Material y método: investigación de tipo descriptiva correlacional. Se examinó a 114 pacientes de ambos sexos, con edades entre 35 y 65 años. Fueron clasificados en dos grupos: un grupo de 38 pacientes sanos y otro de 38 pacientes que tenían perfil lipídico controlado y arte...

  4. Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

    International Nuclear Information System (INIS)

    Bussac, J.; Reuss, P.

    1985-01-01

    This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

  5. Evaluación de bolsa atmósfera modificada y concentraciones de anhídrido sulfuroso aplicadas sobre frutos de arándano alto (Vaccinium corymbosum L. cv. Emerald

    Directory of Open Access Journals (Sweden)

    Mario Rodríguez Beraud

    2015-01-01

    Full Text Available Con el objetivo de evaluar las técnicas de atmósfera modificada y aplicación de anhídrido sulfuroso sobre parámetros de calidad de postcosecha en frutos de arándanos (Vaccinium corymbosum L. cv. Emerald, se realizó un experimento de seis tratamientos, dados por la combinación de dos factores, atmósfera modificada (con y sin, y diferentes concentraciones de anhídrido sulfuroso (generadas por 0, 1 y 2 g de metabisulfito de sodio durante 7, 14, 21 y 28 días a 0 °C. Con la dosis de 2 g de metabisulfito de sodio en atmósfera modificada no se presentaron pudriciones, a diferencia del tratamiento testigo que presentó un 4,86% luego de 28 días de almacenaje. Los resultados indican que la incidencia de pudrición gris disminuyó significativamente (p ≤ 0,05 con anhídrido sulfuroso en bolsa atmósfera modificada, existiendo un efecto de interacción entre ambos factores, no obstante, el gas causó daños de blanqueamiento de frutos, el que correspondió a un 11,66% con una dosis de 2 g de metabisulfito de sodio, luego de 28 días de almacenaje. El uso de bolsa de atmósfera modificada redujo significativamente (p ≤ 0,05 la pérdida de peso por deshidratación (en promedio un 4% respecto a los tratamientos donde esta tecnología no fue utilizada. La concentración de sólidos solubles no fue influenciada por los tratamientos, manteniéndose entre 13 y 14%.

  6. Paleolimnological studies of Laguna Chica of San Pedro (VIII Region: Diatoms, hydrocarbons and fatty acid records Estudio Paleolimnológico de Laguna Chica de San Pedro (VIII Región: Diatomeas, hidrocarburos y ácidos grasos

    Directory of Open Access Journals (Sweden)

    ROBERTO URRUTIA

    2000-12-01

    Full Text Available Diatom, hydrocarbons and fatty acid sedimentary records were used for reconstructing the recent (last 150 years palaeolimnological history of Laguna Chica of San Pedro (Concepción, VIII Región, Chile. Cluster analyses (Constrained Incremental Sum of Squares on the diatom data revealed three distinct periods. The first period (1883-1940's showed a pronounced increase in sedimentation rate and a slight increase in organic matter accumulation. In this period, eutrophic species (Aulacoseira granulata and Staurosira construens became increasingly dominant. From the 1940s until the 1970s the diatom signal is more equivocal: after the initial decrease in the relative abundance of A. granulata and S. construens their numbers fluctuate without a clear pattern. Sedimentation rates strongly fluctuate in this period. From 1978 onwards eutrophic species are in decline while indicators of oligotrophic conditions, such as Cyclotella stelligera and Aulacoseira distans, become more abundant. This shift in the lake trophic status could not be attributed to a reduction in the nutrient load from the catchment and we hypothesize that the invasion of the lake by the submersed macrophyte Egeria densa has altered nutrient availability to the plankton communities. This is in agreement with the hydrocarbons and fatty acid analyses which demonstrate a shift in carbon number distributions from short chain alkanes and alkanoic acids (typical for microalgae to long chain molecules (characteristic for higher plants in the upper layers of the lake sedimentSe realizó la reconstrucción histórica de los últimos 150 años de Laguna Chica de San Pedro (Concepción, VIII Región, Chile, a través de la utilización de los restos de diatomeas, hidrocarburos y ácidos grasos contenidos en la columna de sedimento. El análisis estratigráfico de las diatomeas reveló la presencia de tres períodos diferentes. El primer período (1883-1940's, mostró un marcado aumento de las

  7. TPDWR2: thermal power determination for Westinghouse reactors, Version 2. User's guide

    International Nuclear Information System (INIS)

    Kaczynski, G.M.; Woodruff, R.W.

    1985-12-01

    TPDWR2 is a computer program which was developed to determine the amount of thermal power generated by any Westinghouse nuclear power plant. From system conditions, TPDWR2 calculates enthalpies of water and steam and the power transferred to or from various components in the reactor coolant system and to or from the chemical and volume control system. From these results and assuming that the reactor core is operating at constant power and is at thermal equilibrium, TPDWR2 calculates the thermal power generated by the reactor core. TPDWR2 runs on the IBM PC and XT computers when IBM Personal Computer DOS, Version 2.00 or 2.10, and IBM Personal Computer Basic, Version D2.00 or D2.10, are stored on the same diskette with TPDWR2

  8. Once-through CANDU reactor models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % 235 U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given

  9. Radiation protection at the RA Reactor in 1988, Part -2, RA reactor annual report

    International Nuclear Information System (INIS)

    Ninkovic, M.; Ajdacic, N.; Zaric, M.; Vukovic, Z.

    1988-01-01

    Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor and radiation protection; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Decontamination and relevant actions, collecting and treatment of fluid effluents; and and solid radioactive wastes [sr

  10. Irradiation effects on Zr-2.5Nb in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

  11. Evaluación del osteocoral como material de implante en bolsas infraóseas de dientes multirradiculares

    Directory of Open Access Journals (Sweden)

    Tania Sotomayor Marín

    1999-12-01

    Full Text Available Se evalúa la eficacia del osteocoral como material de implante en el tratamiento de bolsas infraóseas en dientes multirradiculares. Se analizaron 14 pacientes que se dividieron en 2 grupos: el primero incluyó a 6 pacientes con un total de 12 defectos, los cuales se evaluaron hasta los 6 meses. El segundo, con 8 pacientes y 16 defectos, que se reevaluaron a los 12 y 24 meses. En los 2 grupos se incluyeron pacientes de ambos sexos, que fueron implantados con osteocoral (grupo estudio y con hidroxiapatita (grupo control. Se realizó reparación inicial que incluyó remoción de cálculo y pulido de la superficie dentaria, educación y motivación y evaluación del cepillado, que debía mostrar valores iguales o mayores del 80 % en la remoción de placa dentobacteriana. Posteriormente se realizó el implante mediante operación a colgajo. Se realizaron radiografías de control a los 14 días, 6 meses (para el primer grupo y 12 y 24 meses (para el segundo grupo. Se controló sistemáticamente la higiene bucal en ambos grupos. Se controlaron nuevamente los indicadores clínicos a los 6 meses para el primer grupo, y a los 12 y 24 meses para el segundo. Se observó una disminución estadísticamente significativa en el índice gingival, profundidad de la bolsa y movilidad dentaria para ambos materiales implantológicos, sin que se reportaran grandes diferencias entre éstos. Radiográficamente se observó la presencia de relleno en el defecto original, y no hubo reacciones locales adversas, por lo que se consideró efectivo el tratamiento.Effectiveness of osteocoral was assessed as material for implants at infraosseous pockets of multirooted teeth. 14 analised patients were divided into 2 groups: first, included 6 cases and 16 defects, which were evaluated ultil 6 months. Second, included 8 cases and 16 defects, evaluated at 12 and 24 months. In both groups, males and women, were included underwent to implants with osteocoral (study group and

  12. IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-01

    The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1.

  13. IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    1992-01-01

    The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1

  14. Integrando información de carácter temporal y transversal en la predicción del rendimiento inicial de las salidas a bolsa

    Directory of Open Access Journals (Sweden)

    David Quintana Montero

    2007-04-01

    Full Text Available Este artículo aborda el fenómeno del rendimiento inicial de las salidas a bolsa a través de modelos que consideran la cuestión tanto desde un punto de vista longitudinal como transversal. La propuesta consiste en una forma de incorporar tanto la inercia del mercado primario como información relacionada con la estructura de la colocación al estudio de casos concretos. Los resultados ponen de manifiesto una mejora substancial de la capacidad explicativa de las regresiones empleadas.

  15. Characterization of the Three Mile Island Unit-2 reactor building atmosphere prior to the reactor building purge

    International Nuclear Information System (INIS)

    Hartwell, J.K.; Mandler, J.W.; Duce, S.W.; Motes, B.G.

    1981-05-01

    The Three Mile Island Unit-2 reactor building atmosphere was sampled prior to the reactor building purge. Samples of the containment atmosphere were obtained using specialized sampling equipment installed through penetration R-626 at the 358-foot (109-meter) level of the TMI-2 reactor building. The samples were subsequently analyzed for radionuclide concentration and for gaseous molecular components (O 2 , N 2 , etc.) by two independent laboratories at the Idaho National Engineering Laboratory (INEL). The sampling procedures, analysis methods, and results are summarized

  16. Radiation protection at the RA Reactor in 1998, RA reactor annual report, Part -2

    International Nuclear Information System (INIS)

    Ninkovic, M.; Pavlovic, R.; Mandic, M.; Pavlovic, S.; Grsic, Z.

    1998-01-01

    Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Collecting and treatment of fluid effluents; and (4) radioactive wastes, decontamination and actions. Each of the category is described as a separate annex of this report [sr

  17. Active species in a large volume N2-O2 post-discharge reactor

    International Nuclear Information System (INIS)

    Kutasi, K; Pintassilgo, C D; Loureiro, J; Coelho, P J

    2007-01-01

    A large volume post-discharge reactor placed downstream from a flowing N 2 -O 2 microwave discharge is modelled using a three-dimensional hydrodynamic model. The density distributions of the most populated active species present in the reactor-O( 3 P), O 2 (a 1 Δ g ), O 2 (b 1 Σ g + ), NO(X 2 Π), NO(A 2 Σ + ), NO(B 2 Π), NO 2 (X), O 3 , O 2 (X 3 Σ g - ) and N( 4 S)-are calculated and the main source and loss processes for each species are identified for two discharge conditions: (i) p = 2 Torr, f = 2450 MHz, and (ii) p = 8 Torr, f = 915 MHz; in the case of a N 2 -2%O 2 mixture composition and gas flow rate of 2 x 10 3 sccm. The modification of the species relative densities by changing the oxygen percentage in the initial gas mixture composition, in the 0.2%-5% range, are presented. The possible tuning of the species concentrations in the reactor by changing the size of the connecting afterglow tube between the active discharge and the large post-discharge reactor is investigated as well

  18. G 2 reactor project; Projet de pile a double fin: G 2

    Energy Technology Data Exchange (ETDEWEB)

    Ailleret, [Electricite de France (EDF), Dir. General des Etudes de Recherches, 75 - Paris (France); Taranger, P; Yvon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    The CEA actually constructs the G-2 reactor core working with natural uranium, which will use graphite as moderator, and gas under pressure as cooling fluid. This report presents the specificity of the new reactor: - the different elements of the reactor core, - the control and the security of the reactor, - the renewal of the fuel, - the biologic surrounding wall, - and the cooling circuit. (M.B.) [French] le Commissariat a l'Energie Atomique construit actuellement la pile G-2 a Uranium naturel, qui utilisera le graphite comme moderateur, et le gaz sous pression comme fluide de refroidissement. Ce rapport presente les specificite du nouveau reacteur: - les differents elements de la pile, - le controle et la securite du reacteur, - le renouvellement du combustible, - l'enceinte biologique, - et le circuit de refroidissement. (M.B.)

  19. Reactor handbook. 2. rev. ed.

    International Nuclear Information System (INIS)

    Lederer, B.J.; Wildberg, D.W.

    1992-01-01

    On the basis of the guidelines on expert knowledge, the book discusses the subjects of atomic physics, heat transfer, nuclear power plants, reactor materials, radiation protection, reactor safety, reactor instrumentation, and reactor operation, with special regard to nuclear power plants with LWR-type reactors. The book is intended for shift personnel, especially gang bosses, reactor operators, and control station operators: for this reason a practical and rather popular style has been chosen. However, the book will also be a manual for other operating personnel, personnel of producer companies, expert organisations, authorities, and students. It can be used as a textbook for staff training, a manual for the practice, and as accompanying book for teaching at nuclear engineering schools. (orig.) With 173 figs [de

  20. The hamster cheek pouch: an immunologically privileged site suitable to the study of granulomatous infections A bolsa jugal do hamster: um local imunologicamente privilegiado, apropriado para o estudo das ¡nfecções granulomatosas

    Directory of Open Access Journals (Sweden)

    M. S. P. de Arruda

    1995-08-01

    Full Text Available The hamster check pouch is an invagination of oral mucosa, characterized histologically as skin-like. In this paper we describe anatomical, histological and embriological features of the pouch and coment on the pouch as an immunologically privileged site since it lacks lymphatic drainage and has few Langerhans cells. We present the review from literature and our observations after inoculation in the pouch of mycobacteriae (BCG, Mycobacterium tuberculosis and Mycobacterium leprae and a fungus (Paracoccidioides brasiliensis. Lesions in the pouch were granulomatous but smaller and long lasting; even granulomatous, the reaction was inefficient to control the proliferation of agents compared with inoculation in other sites, except for BCG. Appearance of immunity was also delayed or absent and, when it was detected, a sharp decrease in number of agents in pouch lesions was observed. These observations make the pouch an interesting site for the study of the role of immune system in infeccious diseases and in granuloma formation.A bolsa jugal do hamster (BJH é uma invaginação da mucosa oral, caracterizada histologicamente como semelhante a pele. Nesse estudo nós descrevemos algumas de suas características anatômicas, histológicas e embriológicas e comentamos sobre sua propriedade como local imunologicamente privilegiado, considerando a ausência de drenagem linfática e o reduzido número de células de Langerhans. Apresentamos também os resultados obtidos quando da inoculação de micobacterias (BCG, Mycobacterium tuberculosis e Mycobacterium leprae e do fungo Paracoccidioides brasiliensis na bolsa jugal. Comparada com as lesões provocadas em outras localizações e, à exceção do BCG, as lesões induzidas na bolsa são menores e de maior duração e, mesmo quando granulomatosas, incapazes de controlar a multiplicação do agente; nos casos em que houve o desenvolvimento da resposta imune, ele se fez tardiamente e foi acompanhado pela redu

  1. Índice de sostenibilidad empresarial como instrumento en la identificación del compromiso de las sociedades anónimas cotizadas en Bolsa de Valores de Colombia

    OpenAIRE

    Hinestroza Palacio, Santo Alfonso

    2014-01-01

    Tesis (Maestría en Desarrollo Sostenible y Medio Ambiente). Universidad de Manizales. Facultad de Ciencias Contables, Económicas y Administrativas, 2014 Este estudio busco el diseño de un índice de sostenibilidad empresarial aplicable a las empresas colombianas que cotizan en la bolsa de valores de Colombia que tengan un alto compromiso con la responsabilidad social empresarial en el contexto del desarrollo sostenible. Los logros de las empresas en los avances de materia de responsabil...

  2. Annual report on JEN-1 and JEN-2 Reactors; Informe periodico de Reactores JEN-1 y JEN-2 correpondiente al ano 1972

    Energy Technology Data Exchange (ETDEWEB)

    Montes Ponce de Leon, J.

    1974-07-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  3. Trajetórias escolares atípicas : o impacto da bolsa de mérito no projeto de vida dos estudantes

    OpenAIRE

    Santos, Lídia Maria da Silva Calvão Morgado dos

    2012-01-01

    Dissertação de mestrado em Ciências da Educação (área de especialização em Sociologia da Educação e Políticas Educativas) A presente dissertação resultou de uma investigação realizada numa escola secundária pública e incidiu sobre as trajetórias escolares de sucesso de estudantes apoiados pela Ação Social Escolar e beneficiários da Bolsa de Mérito. Trata-se de um estudo descritivo/ interpretativo que aborda a relação entre a atribuição daquele prémio e a construção, por parte dos estudante...

  4. Optimized Control Rods of the BR2 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalcheva, Silva; Koonen, E.

    2007-09-15

    At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

  5. Optimized Control Rods of the BR2 Reactor

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, E.

    2007-01-01

    At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

  6. Upgrading of the research reactors FRG-1 and FRG-2

    International Nuclear Information System (INIS)

    Krull, W.

    1981-01-01

    In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

  7. Uso de osteocoral como material de implante en bolsas infraóseas de dientes Monorradiculares

    Directory of Open Access Journals (Sweden)

    . Yamilé Hernández Alemán,

    1999-12-01

    Full Text Available Se evaluó la eficacia del osteocoral como material de implante en el tratamiento de bolsas infraóseas de dientes monorradiculares. Se realizaron 18 injertos en 17 dientes con defectos angulares, en 6 pacientes de ambos sexos; 9 implantes correspondieron al grupo control con hidroxiapatita y 9 al grupo de estudio que fue implantado con osteocoral. Se realizó preparación inicial que incluyó: remoción de cálculos y pulido de las superficies dentarias, educación y motivación sobre el tratamiento recibido, corrección del cepillado igual o mayor al 80 % en la remoción de placa dentobacteriana. Se realizó el implante mediante operación a colgajo, con sutura y colocación de apósito periodontal. Se realizaron radiografías de control a los 14 días, a los 3 y 6 meses. Se controló sistemáticamente la higiene bucal. A los 6 meses se registraron nuevamente los indicadores clínicos. El análisis final de los resultados mostró una disminución estadísticamente significativa en el índice gingival, profundidad de la bolsa al sondeo y movilidad dentaria para ambos materiales implantológicos. No se reportaron grandes diferencias entre éstos para este tamaño de muestra, no hubo reacciones adversas y se logró la permanencia del implante de osteocoral, por lo que se consideró efectivo el tratamiento.Effectivenes of osteocoral as implant material was assessed to treat infraosseous pockets of multirooted teeth. 18 grafts were inserted in 17 teeth with angular defects in 6 patients of both sexes; 9 implants corresponded to control group (hydroxiapatite and 9 corresponded to study group (osteocoral. Initial preparation included: removal of calculus and polishing of dental surface, education and motivation about treatment applied, correction of tooth-brushing equal or greater 80 % in removal of dentobacterial plaque. Implant was inserted by flap surgery using suture and placement of periodontal dresssing. Control X-rays were made within 14 days

  8. Impact of Bolsa Família Program on the nutritional status of children and adolescents from two Brazilian regions

    Directory of Open Access Journals (Sweden)

    Naiara SPERANDIO

    Full Text Available ABSTRACT Objective: To assess and compare the impact of the Bolsa Família Program (Family Allowance on the nutritional status of children and adolescents from the Brazilian Northeastern and Southeastern regions. Methods: The study used data from a database derived from a subsample of the Family Budget Survey conducted from 2008 to 2009. The ratios of underweight, stunted, and overweight children were calculated. Impact measurement analysis was preceded by propensity score matching, which matches beneficiary and non-beneficiary families in relation to a set of socioeconomic features. The nearest-neighbor matching algorithm estimated the program impact. Results: The ratio of underweight children and adolescents was, on average, 1.1% smaller in the beneficiary families than in the non-beneficiary families in the Northeastern region. As for the Southeastern region, the ratio of overweight children and adolescents was, on average, 4.2% smaller in the beneficiary families. The program did not affect stunting in either region. Conclusion: The results showed the positive impact and good focus of the program. Thus, once linked to structural actions, the program may help to improve the nutritional status and quality of life of its beneficiaries.

  9. Annual report on JEN-1 and JEN-2 Reactors

    International Nuclear Information System (INIS)

    Montes Ponce de Leon, J.

    1974-01-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  10. System Design of a Supercritical CO_2 cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Kim, Seong Gu; Cho, Seongkuk; Yu, Hwanyeal; Kim, Yonghee; Jeong, Yong Hoon; Lee, Jeong Ik

    2014-01-01

    Small modular reactor (SMR) systems that have advantages of little initial capital cost and small restriction on construction site are being developed by many research organizations around the world. Existing SMR concepts have the same objective: to achieve compact size and a long life core. Most of small modular reactors have much smaller size than the large nuclear power plant. However, existing SMR concepts are not fully modularized. This paper suggests a complete modular reactor with an innovative concept for reactor cooling by using a supercritical carbon dioxide. The authors propose the supercritical CO_2 Brayton cycle (S-CO_2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the reactor core and PCU in one vessel. A conceptual design of the proposed small modular reactor was developed, which is named as KAIST Micro Modular Reactor (MMR). The supercritical CO_2 Brayton cycle for the S-CO_2 cooled reactor core was optimized and the size of turbomachinery and heat exchanger were estimated preliminary. The nuclear fuel composed with UN was proposed and the core lifetime was obtained from a burnup versus reactivity calculation. Furthermore, a system layout with fully passive safety systems for both normal operation and emergency operation was proposed. (author)

  11. The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

    International Nuclear Information System (INIS)

    Townes, B.M.; Hilborn, J.W.

    1985-06-01

    A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

  12. Operation of the SLOWPOKE-2 reactor in Jamaica

    Energy Technology Data Exchange (ETDEWEB)

    Grant, C.N.; Lalor, G.C.; Vuchkov, M.K. [University of the West Indies, Kingston (Jamaica)

    2001-07-01

    Over the past sixteen years lCENS has operated a SLOWPOKE 2 nuclear reactor almost exclusively for the purpose of neutron activation analysis. During this period we have adopted a strategy of minimum irradiation times while optimizing our output in an effort to increase the lifetime of the reactor core and to maintaining fuel integrity. An inter-comparison study with results obtained with a much larger reactor at IPEN has validated this approach. The parameters routinely monitored at ICENS are also discussed and the method used to predict the next shim adjustment. (author)

  13. Irradiation techniques at BR2 reactor

    International Nuclear Information System (INIS)

    Hebel, W.

    1978-01-01

    Since 1963 the material testing reactor BR2 at Mol is operated for the realisation of numerous research programs and experiments on the behavior of materials under nuclear radiation and in particular under intensive neutron exposure. During this period special irradiation techniques and experimental devices were developed according to the desiderata of the different experiments and to the irradiation possibilities offered at BR2. The design and the operating characteristics of quite a number of those irradiation rigs of proven reliability may be used or can be made available for new irradiation experiments. A brief description is given of some typical irradiation devices designed and constructed by CEN/SCK, Technology and Energy Dpt. They are compiled according to their main use for the different research and development programs realized at BR2. Their eventual application however for different objectives could be possible. A final chapter summarizes the principal irradiation conditions offered by BR2 reactor. (author)

  14. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1985-09-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  15. TMI-2 reactor vessel plenum final lift

    International Nuclear Information System (INIS)

    Wilson, D.C.

    1986-01-01

    Removal of the plenum assembly from the TMI-2 reactor vessel was necessary to gain access to the core region for defueling. The plenum was lifted from the reactor vessel by the polar crane using three specially designed pendant assemblies. It was then transferred in air to the flooded deep end of the refueling canal and lowered onto a storage stand where it will remain throughout the defueling effort. The lift and transfer were successfully accomplished on May 15, 1985 in just under three hours by a lift team located in a shielded area within the reactor building. The success of the program is attributed to extensive mockup and training activities plus thorough preparations to address potential problems. 54 refs

  16. Refurbishment programme for the BR2-reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koonen, E [Centre d' Etude de l' Energie Nucleaire, Studiecentrum voor Kernenergie, BR2 Department, Boeretang, Mol (Belgium)

    1992-07-01

    BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

  17. Refurbishment programme for the BR2-reactor

    International Nuclear Information System (INIS)

    Koonen, E.

    1992-01-01

    BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

  18. Refurbishing the BR2 materials testing reactor

    International Nuclear Information System (INIS)

    Baugnet, J.M.; Dekeyser, J.; Gubel, P.

    1995-01-01

    SCK/CEN is refurbishing its BR2 reactor to allow its further operation during the next 15 years; in doing so, it chooses to keep BR2 available for future scientific and technological irradiation programs within an international context. (author) 2 figs

  19. Proceedings of 2. Yugoslav symposium on reactor physics, Part 2, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 2 of the Proceedings of 2. Yugoslav symposium on reactor physics includes eight papers dealing with the following topics: method for measuring high anti reactivities of a reactor system; integration method for thermal reaction rate calculation; Determination of initial core configuration for BHWR-200 MWe; safety shutdowns and failures of the RA reactor equipment; determining the reactivity of absorption rods; measurements of thermal and fast neutron fluxes at the TRIGA reactor and other measurements during operation of the TRIGA reactor; mathematical modelling of the reactor safety; review of problems and methods for radiation risk assessment in the environment of a nuclear power plant

  20. SIRIUS 2: A versatile medium power research reactor

    International Nuclear Information System (INIS)

    Rousselle, P.

    1992-01-01

    Most of the Research Reactors in the world have been critical in the Sixties and operated for twenty to thirty years. Some of them have been completely shut down, modified, or simply refurbished; the total number of RR in operation has decreased but there is still an important need for medium power research reactors in order: - to sustain a power program with fuel and material testing for NPP or fusion reactors; - to produce radioisotopes for industrial or medical purposes, doped silicon, NAA or neutron radiography; - to investigate further the condensed matter, with cold neutrons routed through neutron guides to improved equipment; - to develop new technologies and applications such as medical alphatherapy. Hence, taking advantage of nearly hundred reactor x years operation and backed up by the CEA experience, TECHNICATOME assisted by FRAMATOME has designed a new versatile multipurpose Research Reactor (20-30 Mw) SIRIUS 2 taking into account: - more stringent safety rules; - the lifetime; - the flexibility enabling a wide range of experiments and, - the future dismantling of the facility according to the ALARA criteria

  1. OTUS - Reactor inventory management system based on ORIGEN2

    Energy Technology Data Exchange (ETDEWEB)

    Poellaenen, R; Toivonen, H; Lahtinen, J; Ilander, T

    1995-10-01

    ORIGEN2 is a computer code that calculates nuclide composition and other characteristics of nuclear fuel. The use of ORIGEN2 requires good knowledge in reactor physics. However, once the input has been defined for a particular reactor type, the calculations can be easily repeated for any burnup and decay time. This procedure produces large output files that are difficult to handle manually. A new computer code, known as OTUS, was designed to facilitate the postprocessing of the data. OTUS makes use of the inventory files precalculated with ORIGEN2 in a way that enables their versatile treatment for different safety analysis purposes. A data base is created containing a comprehensive set of ORIGEN2 calculations as a function of fuel burnup and decay time. OTUS is a reactor inventory management system for a microcomputer with Windows interface. Four major data operations are available: (1) Build data modifies ORIGEN2 output data into a suitable format, (2) View data enables flexible presentation of the data as such, (3) Different calculations, such as nuclide ratios and hot particle characteristics, can be performed for severe accident analyses, consequence analyses and research purposes, (4) Summary files contain both burnup dependent and decay time dependent inventory information related to the nuclide and the reactor specified. These files can be used for safeguards, radiation monitoring and safety assessment. (orig.) (22 refs., 29 figs.).

  2. The 2nd reactor core of the NS Otto Hahn

    International Nuclear Information System (INIS)

    Manthey, H.J.; Kracht, H.

    1979-01-01

    Details of the design of the 2nd reactor core are given, followed by a brief report summarising the operating experience gained with this 2nd core, as well as by an evaluation of measured data and statements concerning the usefulness of the knowledge gained for the development of future reactor cores. Quite a number of these data have been used to improve the concept and thus the specifications for the fuel elements of the 3rd core of the reactor of the NS Otto Hahn. (orig./HP) [de

  3. EL-2 reactor: Thermal neutron flux distribution

    International Nuclear Information System (INIS)

    Rousseau, A.; Genthon, J.P.

    1958-01-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  4. Physics design of advanced steady-state tokamak reactor A-SSTR2

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Ushigusa, Kenkichi

    2000-10-01

    Based on design studies on the fusion power reactor such as the DEMO reactor SSTR, the compact power reactor A-SSTR and the DREAM reactor with a high environmental safety and high availability, a new concept of compact and economic fusion power reactor (A-SSTR2) with high safety and high availability is proposed. Employing high temperature superconductor, the toroidal filed coils supplies the maximum field of 23T on conductor which corresponds to 11T at the magnetic axis. A-SSTR2 (R p =6.2m, a p =1.5m, I p =12MA) has a fusion power of 4GW with β N =4. For an easy maintenance and for an enough support against a strong electromagnetic force on coils, a poloidal coils system has no center solenoid coils and consists of 6 coils located on top and bottom of the machine. Physics studies on the plasma equilibrium, controllability of the configuration, the plasma initiation and non-inductive current ramp-up, fusion power controllability and the diverter have shown the validity of the A-SSTR2 concept. (author)

  5. FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients

    International Nuclear Information System (INIS)

    Fox, J.N.; Lawler, B.E.; Butz, H.R.; Heames, T.J.

    1984-01-01

    1 - Description of problem or function: FORE2 is a coupled thermal hydraulics-point kinetics digital computer code designed to calculate significant reactor parameters under steady-state conditions, or as functions of time during transients. The transients may result from a programmed reactivity insertion or a power change. Variable inlet coolant flow rate and temperature are considered. The code calculates the reactor power, the individual reactivity feedbacks, and the temperature of coolant, cladding, fuel, structure, and additional material for up to seven axial positions in three channel types which represent radial zones of the reactor. The heat of fusion, accompanying fuel melting, the liquid metal voiding reactivity, and the spatial and the time variation of the fuel cladding gap coefficient due to changes in gap size are considered. 2 - Method of solution: FORE2 input consists of property data, geometry, power and flow distribution factors, external time varying functions, experimental coefficients, and termination data. The differential equations for fluid flow, heat transfer, and point neutronics are solved by explicit finite-difference procedures. 3 - Restrictions on the complexity of the problem: Reactor excursions which can be calculated are restricted to those transients in which the reactor is not substantially destroyed. As a general rule, changes in reactor geometry and composition during an excursion are limited to those cases in which the reactivity effects of the changes may be considered as small perturbations of the initial system. Thus, accidents involving large-scale disassembly and bulk meltdown of a core are not covered by FORE2. FORE2 is valid only while the core retains its initial geometry

  6. Avaliação antropométrica e consumo alimentar em crianças menores de cinco anos residentes em um município da região do semiárido nordestino com cobertura parcial do programa bolsa família Anthropometric assessment and food intake of children younger than 5 years of age from a city in the semi-arid area of the Northeastern region of Brazil partially covered by the bolsa família program

    Directory of Open Access Journals (Sweden)

    Silvia Regina Dias Médici Saldiva

    2010-04-01

    Full Text Available OBJETIVO: Avaliar as condições de saúde e nutrição de crianças menores de cinco anos, e associar a qualidade do consumo alimentar aos beneficiários do Programa Bolsa Família de um município do semiárido brasileiro. MÉTODOS: Foram avaliadas 189 crianças, a partir de uma amostragem de 411 domicílios do município de João Câmara (RN. Foram realizadas medidas de peso e altura, e levantadas às condições socioeconômicas e determinação dos hábitos alimentares. Para o diagnóstico nutricional das crianças foram utilizados os indicadores Peso/Idade, Altura/Idade e Peso/Altura. Análises univariadas foram realizadas e modelos bivariados e multivariados de regressão logística foram construídos para testar a hipótese do estudo. RESULTADOS: O déficit de peso foi de 4,3% e o de altura de 9,9%, e o excesso de peso de 14,0%. Não foram encontradas diferenças estatísticas entre o estado nutricional de crianças beneficiárias e não beneficiárias do Programa Bolsa Família. Em ambos os grupos, os consumos de frutas, verduras e legumes foram baixos e semelhantes entre si. As crianças do programa bolsa família têm risco três vezes maior de consumir guloseimas (OR 3,06 - IC 1,35-6,95. CONCLUSÃO: Os resultados do padrão de consumo alimentar dessa população apontam para uma situação de "risco alimentar e nutricional", e exigem uma intervenção por parte dos profissionais de saúde para a promoção da alimentação saudável.OBJECTIVE: The objective of this study was to assess the health and nutritional status of children under five years of age and to associate the quality of the foods consumed with the Bolsa Família Program in a city located in the Brazilian semi-arid region. METHOD: A total of 189 children from a sample of 411 households in the city of João Câmara (RN were assessed. Weight and height were measured and socioeconomic and food habits were determined with the use of questionnaires. The nutritional status of

  7. Keeping research reactors relevant: A pro-active approach for SLOWPOKE-2

    International Nuclear Information System (INIS)

    Cosby, L.R.; Bennett, L.G.I.; Nielsen, K.; Weir, R.

    2010-01-01

    The SLOWPOKE is a small, inherently safe, pool-type research reactor that was engineered and marketed by Atomic Energy of Canada Limited (AECL) in the 1970s and 80s. The original reactor, SLOWPOKE-1, was moved from Chalk River to the University of Toronto in 1970 and was operated until upgraded to the SLOWPOKE-2 reactor in 1973. In all, eight reactors in the two versions were produced and five are still in operation today, three having been decommissioned. All of the remaining reactors are designated as SLOWPOKE-2 reactors. These research reactors are prone to two major issues: aging components and lack of relevance to a younger audience. In order to combat these problems, one SLOWPOKE -2 facility has embraced a strategy that involves modernizing their reactor in order to keep the reactor up to date and relevant. In 2001, this facility replaced its aging analogue reactor control system with a digital control system. The system was successfully commissioned and has provided a renewed platform for student learning and research. The digital control system provides a better interface and allows flexibility in data storage and retrieval that was never possible with the analogue control system. This facility has started work on another upgrade to the digital control and instrumentation system that will be installed in 2010. The upgrade includes new computer hardware, updated software and a web-based simulation and training system that will allow licensed operators, students and researchers to use an online simulation tool for training, education and research. The tool consists of: 1) A dynamic simulation for reactor kinetics (e.g., core flux, power, core temperatures, etc). This tool is useful for operator training and student education; 2) Dynamic mapping of the reactor and pool container gamma and neutron fluxes as well as the vertical neutron beam tube flux. This research planning tool is used for various researchers who wish to do irradiations (e.g., neutron

  8. Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

    International Nuclear Information System (INIS)

    Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

    1983-02-01

    The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems

  9. Thermal reactor benchmark tests on JENDL-2

    International Nuclear Information System (INIS)

    Takano, Hideki; Tsuchihashi, Keichiro; Yamane, Tsuyoshi; Akino, Fujiyoshi; Ishiguro, Yukio; Ido, Masaru.

    1983-11-01

    A group constant library for the thermal reactor standard nuclear design code system SRAC was produced by using the evaluated nuclear data JENDL-2. Furthermore, the group constants for 235 U were calculated also from ENDF/B-V. Thermal reactor benchmark calculations were performed using the produced group constant library. The selected benchmark cores are two water-moderated lattices (TRX-1 and 2), two heavy water-moderated cores (DCA and ETA-1), two graphite-moderated cores (SHE-8 and 13) and eight critical experiments for critical safety. The effective multiplication factors and lattice cell parameters were calculated and compared with the experimental values. The results are summarized as follows. (1) Effective multiplication factors: The results by JENDL-2 are considerably improved in comparison with ones by ENDF/B-IV. The best agreement is obtained by using JENDL-2 and ENDF/B-V (only 235 U) data. (2) Lattice cell parameters: For the rho 28 (the ratio of epithermal to thermal 238 U captures) and C* (the ratio of 238 U captures to 235 U fissions), the values calculated by JENDL-2 are in good agreement with the experimental values. The rho 28 (the ratio of 238 U to 235 U fissions) are overestimated as found also for the fast reactor benchmarks. The rho 02 (the ratio of epithermal to thermal 232 Th captures) calculated by JENDL-2 or ENDF/B-IV are considerably underestimated. The functions of the SRAC system have been continued to be extended according to the needs of its users. A brief description will be given, in Appendix B, to the extended parts of the SRAC system together with the input specification. (author)

  10. WWER-1000 reactor simulator. Material for training courses and workshops. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No.12, Reactor Simulator Development (2001). Course material for workshops using a pressurized water reactor (PWR) simulator developed for the IAEA by Cassiopeia Technologies Inc. of Canada is presented in the IAEA publication, Training Course Series No. 22, 2nd edition, Pressurized Water Reactor Simulator (2005) and Training Course Series No.23, 2nd edition, Boiling Water Reactor Simulator (2005). This report consists of course material for workshops using the WWER-1000 Reactor Department Simulator from the Moscow Engineering and Physics Institute, Russian Federation

  11. ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

    International Nuclear Information System (INIS)

    Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

    2010-01-01

    The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

  12. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 2-2

    International Nuclear Information System (INIS)

    Okuda, Eiji; Ito, Hiromichi; Yoshihara, Shizuya

    2014-01-01

    An accident occurred in experimental fast reactor 'Joyo' in 2007 which is obstruction of fuel change equipment caused by contacting rotating plug and MARICO-2. In addition, we confirmed two happenings in the reactor vessel that (1) Deformation of MARICO-2 subassembly on the in vessel storage rack together with a transfer pot, (2) Deformation of the Upper core structure of 'Joyo' caused by contacting MARICO-2 subassembly and the UCS. We do the restoration work for restoring it. This time, we describe current status of Replacement work of the UCS. (author)

  13. Estudio de la biodegradación de bolsas oxo - biodegradables utilizando compost maduro seco, con aireación y simulando condiciones ambientales de humedad y temperatura de un relleno sanitario ubicado en la Costa Ecuatoriana.

    OpenAIRE

    Sandoval Moreira, María Isabel

    2014-01-01

    This work allowed the study of biodegradability of four types of oxo-biodegradable bags used to sell products, simulating environmental conditions of a landfill located in the city of Manta. El presente trabajo permitió el estudio de la biodegrabilidad de cuatro tipos de bolsas oxo-biodegradables utilizadas para la venta de productos, simulando condiciones ambientales de un relleno sanitario ubicado en la ciudad de Manta.

  14. Um debate tridimensional sobre os padrões de proteção social no Brasil frente à crise capitalista internacional : o caso do Bolsa Família

    OpenAIRE

    Maurício, Márcio Fernandes

    2011-01-01

    Esta dissertação propõe um debate teórico sobre os padrões de proteção social no Brasil frente à crise capitalista internacional, com um estudo de caso sobre o Programa Bolsa Família (PBF). O recorte de análise consiste na articulação do PBF com os chamados "programas complementares", no âmbito da coordenação intergovernamental e interdependência organizacional, preconizadas na Constituição Federal de 1988. E, o objeto de análise, neste caso, corresponde ao debate teórico propriamente dito. A...

  15. Sterilization of E. coli bacterium in a flowing N2-O2 post-discharge reactor

    International Nuclear Information System (INIS)

    Villeger, S; Cousty, S; Ricard, A; Sixou, M

    2003-01-01

    Effective destruction of Escherichia coli (E. coli) bacteria has been obtained in a flowing N 2 -O 2 microwave post-discharge reactor. The sterilizing agents are the O atoms and the UV emissions of NOβ which are produced by N and O atoms recombination in the reactor. In the following plasma conditions: pressure 5 Torr, flow rate 1 L n min -1 , microwave power of 100 W in a quartz tube of 5 mm, an O atom density of 2.5x10 15 cm -3 is measured by NO titration in the post-discharge reactor with UV emission in a N 2 -(5%)O 2 gas mixture. Full destruction of 10 13 cfu ml -1 E. coli is observed after a treatment time of 25 min. (rapid communication)

  16. Reptile and rodent parasites in raptor pellets in an archaeological context: the case of Epullán Chica (northwestern Patagonia, Argentina)

    Science.gov (United States)

    Beltrame, María Ornela; Fernández, Fernando Julián; Sardella, Norma Haydeé

    2015-07-01

    Paleoparasitology is the study of parasite remains from archaeological and paleontological sites. Raptor pellets can be used as source for paleoparasitological information in archaeological sites. However, this zooarchaeological material has been scarcely studied. Epullán Chica (ECh) is an archaeological site in northwestern Patagonia. This cave yielded remains from more than 2000 years before present. The aim of this paper was to study the parasite remains found in owl pellets from the archaeological site ECh, and to discuss the paleoparasitological findings in an archaeological context. Twenty two raptor pellets were examined for parasites. The pellets were whole processed by rehydration in a 0.5% water solution of trisodium phosphate, followed by homogenization, filtered and processed by spontaneous sedimentation. Eight out of 22 bird pellets examined were positive for parasites from reptiles and rodents. Representatives of 12 parasite taxa were recorded; nine of this parasitic species were reported for the first time from ancient samples from Patagonia. This is the first time that pellets give evidences of ancient reptile parasites from archaeological contexts. It is noteworthy that Late Holocene hunter-gatherers of the upper Limay River basin, could have been exposed to some of these zoonotic parasites. Future paleoparasitological studies on owl pellets may reflect even more the parasitological diversity of all micromammal and reptile species presents in ancient times.

  17. Safe dismantling of the SVAFO research reactors R2 and R2-0 in Sweden

    International Nuclear Information System (INIS)

    ARNOLD, Hans-Uwe; BROY, Yvonne; Dirk Schneider

    2017-01-01

    The R2 and R2-0 reactors were part of the Swedish government's research program on nuclear power from the early 1960's. Both reactors were shut down in 2005 following a decision by former operator Studsvik Nuclear AB. The decommissioning of the R2 and R2-0 reactors is divided into three phases. The first phase - awarded to AREVA - involved dismantling of the reactors and associated systems in the reactor pool, treatment of the disassembled components as well as draining, cleaning and emptying the pool. In the second phase, the pool structure itself will be dismantled, while removal of remaining reactor systems, treatment and disposal of materials and clean-up will be carried out in the third stage. The entire work is planned to be completed before the end of this decade. The paper describes the several steps of phase 1 - starting with the team building, followed by the dismantling operations and covers challenges encountered and lessons learned as well. The reactors consist of 5.400 kg aluminum, 6.000 kg stainless steel restraint structures as well as, connection elements of the mostly flanged components (1.000 kg). The most demanding - from a radiological point of view - was the R2-0 reactor that was limited to ∼ 1 m"3 construction volumes but with an extremely heterogeneous activation profile. Based on the calculated radiological entrance data and later sampling, nuclide vectors for both reactors depending on the real placement of the single component and on the material (aluminum and stainless steel) were created. Finally, for the highest activated component from R2 reactor, 85 Sv/h were measured. The dismantling principles - adopted on a safety point of view - were the following: The always protected base area of the ponds served as a flexible buffer area for waste components and packaging. Specific protections were also installed on the walls to protect them from mechanical stress which may occur during dismantling work. A specific work platform was

  18. Benchmark calculations for VENUS-2 MOX -fueled reactor dosimetry

    International Nuclear Information System (INIS)

    Kim, Jong Kung; Kim, Hong Chul; Shin, Chang Ho; Han, Chi Young; Na, Byung Chan

    2004-01-01

    As a part of a Nuclear Energy Agency (NEA) Project, it was pursued the benchmark for dosimetry calculation of the VENUS-2 MOX-fueled reactor. In this benchmark, the goal is to test the current state-of-the-art computational methods of calculating neutron flux to reactor components against the measured data of the VENUS-2 MOX-fuelled critical experiments. The measured data to be used for this benchmark are the equivalent fission fluxes which are the reaction rates divided by the U 235 fission spectrum averaged cross-section of the corresponding dosimeter. The present benchmark is, therefore, defined to calculate reaction rates and corresponding equivalent fission fluxes measured on the core-mid plane at specific positions outside the core of the VENUS-2 MOX-fuelled reactor. This is a follow-up exercise to the previously completed UO 2 -fuelled VENUS-1 two-dimensional and VENUS-3 three-dimensional exercises. The use of MOX fuel in LWRs presents different neutron characteristics and this is the main interest of the current benchmark compared to the previous ones

  19. PCU arrangement of a supercritical CO{sub 2} cooled micro modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong Gu; Baik, Seungjoon; Cho, Seong Kuk; Oh, Bong Seong; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    As part of the SMR(Small Modular Reactor)s development effort, the authors propose a concept of supercritical CO{sub 2} (S-CO{sub 2}) cooled fast reactor combined with the S-CO{sub 2} Brayton cycle. The reactor concept is named as KAIST Micro Modular Reactor (MMR). The S-CO{sub 2} Brayton cycle has many strong points when it is used for SMR's power conversion unit. It occupies small footprints due to the compact cycle components and simple layout. Thus, a concept of one module containing the S-CO{sub 2} cooled fast reactor and power conversion system is possible. This module can be shipped via ground transportation (by trailer) or marine transportation. In this study, the authors propose a new conceptual layout for the S-CO{sub 2} cooled direct cycle while considering various issues for arranging cycle components. The new design has an improved cycle efficiency (from 31% to 34%) than the earlier version of MMR by reducing pressure drops in the heat exchangers. As a more efficient option, a recompression recuperated cycle was also designed. It improves 5% of thermal efficiency while 18tons of mass can be added in comparison to the simple recuperated cycle. Even if we adopt recompression cycle as a PCU, the weight of module (152tons) is less than the ground transportable limit (260tons)

  20. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1984-12-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  1. Neutronic study using oxide and nitride fuels for the Super Phenix 2 reactor

    International Nuclear Information System (INIS)

    Batista, J.L.; Renke, C.A.C.

    1991-11-01

    This report presents a neutronic analysis and a description of the Super Phenix 2 reactor, taken as reference. We present the methodology and results for cell and global reactor calculations for oxide (U O 2 - Pu O 2 ) and nitride (U N - Pu N) fuels. To conclude we compare the performance of oxide and nitride fuels for the reference reactor. (author)

  2. Estereotipos de género 2.0: Auto-representaciones de adolescentes en Facebook

    Directory of Open Access Journals (Sweden)

    2016-07-01

    Full Text Available Chicas y chicos adolescentes hacen un uso diferente de las redes sociales online, y las chicas presentan un mayor riesgo de verse perjudicadas por un uso no adaptativo. El objetivo de este estudio era investigar en qué medida los adolescentes se presentan en términos de estereotipos de género en sus perfiles de Facebook. Los participantes, 623 usuarios de Facebook de ambos sexos, contestaron el Bem Sex Role Inventory (BSRI y el Personal Well-being Index (PWI. En la primera fase, respondieron sobre cómo ven a un adulto típico en términos de estereotipos de género. En la segunda fase, la mitad de ellos contestó el BSRI en relación a cómo se ven a sí mismos, y la otra mitad cómo se presentan en Facebook. Los resultados muestran que los adolescentes se consideran más sexualmente indiferenciados que un adulto típico de su mismo sexo, tanto en su auto-percepción como en su presentación en Facebook. Se confirma que el bienestar psicológico de las chicas baja considerablemente con la edad, y que está asociado a un mayor grado de masculinidad. Se concluye que los adolescentes producen representaciones verdaderas en sus perfiles de Facebook, y que existe una tendencia hacia una auto-concepción y auto-presentación más sexualmente indiferenciada con una leve preferencia por rasgos masculinos, tanto en chicos como en chicas; además, la masculinidad está asociada a un mayor grado de bienestar psicológico.

  3. Shadow corrosion evaluation in the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Sanders, Ch.; Lysell, G.

    2000-01-01

    Post-irradiation examination has shown that increased corrosion occurs when zirconium alloys are in contact with or in proximity to other metallic objects. The observations indicate an influence of irradiation from the adjacent component as the enhanced corrosion occurs as a 'shadow' of the metallic object on the zirconium surface. This phenomenon could ultimately limit the lifetime of certain zirconium alloy components in the reactor. The Studsvik R2 materials test reactor has an In-Core Autoclave (INCA) test facility especially designed for water chemistry and materials research. The INCA facility has been evaluated and found suitable for shadow corrosion studies. The R2 reactor core containing the INCA facility was modeled with the Monte Carlo N-Particle (MCNP) code in order to evaluate the electron deposition in various materials and to develop a hypothesis of the shadow corrosion mechanism. (authors)

  4. Benchmark tests of JENDL-3.2 for thermal and fast reactors

    International Nuclear Information System (INIS)

    Takano, Hideki

    1995-01-01

    Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

  5. Radionuclide distribution in TMI-2 reactor building basement liquids and solids

    International Nuclear Information System (INIS)

    Horan, J.T.; McIsaac, C.V.; Keefer, D.G.

    1984-01-01

    As a result of the TMI-2 accident, approximately 2.46 x 10 6 L of contaminated water were released to the Reactor Building basement. The principal fission product release pathway from the damaged core was through the reactor coolant system (RCS) to the pressurizer, through the pressure-operated relief valve (PORV) on the pressurizer to the Reactor Coolant Drain Tank (RCDT), and then through the RCDT rupture disk to the Reactor Building basement. Since August 1979, a number of efforts have been made to determine the location, quantity, and composition of fission products released to the Reactor Building basement. These efforts have included sampling of the basement water and solids, the basement sump pump recirculation line, the RCDT, and visual surveys using a closed circuit television (CCTV) system. The analysis of basement samples has provided data on the physical and radioisotopic characteristics of the liquids and solids. This paper describes the sample collection techniques and discusses radiochemical analyses results

  6. An experimental investigation of fission product release in SLOWPOKE-2 reactors

    International Nuclear Information System (INIS)

    Harnden, A.M.C.

    1995-09-01

    Increasing radiation fields due to a release of fission products in the reactor container of several SLOWPOKE-2 reactors fuelled with a highly-enriched uranium (HEU) alloy core have been observed. It is believed that these increases are associated with the fuel fabrication where a small amount of uranium-bearing material is exposed to the coolant at the end-welds of the fuel element. To investigate this phenomenon samples of reactor water and gas from the headspace above the water have been obtained and examined by gamma spectrometry methods for reactors of various burnups at the University of Toronto, Ecole Polytechnique and Kanata Isotope Production Facility. An underwater visual examination of the fuel core at Ecole Polytechnique has also provided information on the condition of the core. This report (Volume 1) summarizes the equipment, analysis techniques and results of tests conducted at the various reactor sites. The data report is published as Volume 2. (author). 30 refs., 9 tabs., 20 figs

  7. Estudo físico e físico-químico de diferentes filmes de bolsas de sangue visando a segurança frente ao processamento hemoterapêutico Physical and physicochemical study of different blood bag films in respect to safety during hemotherapeutic processing

    Directory of Open Access Journals (Sweden)

    Armando V. Verceze

    2006-06-01

    Full Text Available Muitas rupturas de bolsas de sangue no processamento e armazenamento levam à abertura do sistema e à perda do conteúdo, com prejuízos econômicos, riscos biológicos e aspectos sociais pela doação voluntária (dados levantados junto a serviços de hemoterapia pelo autor. O propósito foi avaliar "in vitro", por meio de teste cego, diferentes filmes de bolsas de poli (cloreto de vinila-PVC para coleta de sangue disponíveis no mercado nacional, sendo três produzidas no Brasil e duas no exterior, utilizando parâmetros físico e físico-químico. Estas bolsas possuem características especiais como: composição química conforme a Farmacopéia Européia, flexibilidade para enchimento com sangue e resistência a diferentes condições de temperatura e tempo de centrifugação. A fabricação das bolsas ocorre por soldagem por radiofreqüência. A área definida de solda ou costura entre os filmes tem sido apontada como o principal ponto vulnerável a micro-rupturas, durante a centrifugação. Os parâmetros estudados foram: absorção no infravermelho (IR-FT e análise mecânica de tensão-elongação/ruptura, realizados no corpo da bolsa e na solda ou costura. Os espectros (IR-FT foram semelhantes, porém diferentes resultados foram observados na análise mecânica quando comparados entre si. Evidenciamos dois grupos de comportamentos quanto à concentração de grupamentos químicos no infravermelho. Não obtivemos informações da concentração química, do processamento e possíveis diferenças de técnicas empregadas. Os resultados nos permitem concluir que existem diferenças entre as cinco bolsas. Estas propriedades são tão importantes quanto as características biológicas ou bioquímicas. Não encontramos na literatura valores que possam caracterizar qual bolsa seria mais ou menos eficiente frente ao processamento ao qual são submetidas em toda sua cadeia desde a indústria até a transfusão.Many ruptures of blood bags used

  8. The Oak Ridge Research Reactor: safety analysis: Volume 2, supplement 2

    International Nuclear Information System (INIS)

    Hurt, S.S.

    1986-11-01

    The Oak Ridge Research Reactor Safety Analysis was last updated via ORNL-4169, Vol. 2, Supplement 1, in May of 1978. Since that date, several changes have been effected through the change-memo system described below. While these changes have involved the cooling system, the electrical system, and the reactor instrumentation and controls, they have not, for the most part, presented new or unreviewed safety questions. However, some of the changes have been based on questions or recommendations stemming from safety reviews or from reactor events at other sites. This paper discusses those changes which were judged to be safety related and which include revisions to the syphon-break system and changes related to seismic considerations which were very recently completed. The maximum hypothetical accident postulated in the original safety analysis requires dynamic containment and filtered flow for compliance with 10CFR100 limits at the site boundary

  9. Loss of coolant analysis for the tower shielding reactor 2

    International Nuclear Information System (INIS)

    Radcliff, T.D.; Williams, P.T.

    1990-06-01

    The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

  10. Dalhousie SLOWPOKE-2 reactor: A nuclear analytical chemistry facility

    International Nuclear Information System (INIS)

    Chatt, A.; Holzbecher, J.

    1990-01-01

    SLOWPOKE is an acronym for Safe Low POwer Kritical Experiment. The SOWPOKE-2 is a compact, inherently safe, swimming-pool-type reactor designed by the Atomic Energy of Canada Limited for neutron activation analysis (NAA) and isotope production. The Dalhousie University SLOWPOKE-2 reactor (DUSR) has been operating since 1976; a large beryllium reflector was added in 1986 to extend its lifetime by another 8 to 10 yr. The DUSR is generally operated at half-power with a maximum thermal flux of 1.1 x 10 12 n/cm 2 ·s in the inner pneumatic sites and that of 5.4 x 10 11 n/cm 2 ·s in the outer sites. Despite this comparatively low flux, SLOWPOKE-2 reactors have many beneficial features that are continuously being exploited at the DUSR facility for developing nuclear analytical methods for fundamental as well as applied studies. Although NAA is a well-established analytical technique, much of the activation analysis being performed in most facilities has been limited to methods using fairly long-lived nuclides. The approach at the DUSR facility has been to utilize the highly homogeneous, stable, and reproducible neutron flux to develop NAA methods based on short-lived nuclides. SLOWPOKE reactors have a fairly high epithermal neutron flux, which is being advantageously used for determining several trace elements in complex matrices. Radiochemical NAA (RNAA) methods using coprecipitation, distillation, and ion-exchange separations have been used for the determination of very low levels of several elements in biological materials

  11. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    International Nuclear Information System (INIS)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B.

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative

  12. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  13. Dynamic simulation of the 2 MWt slowpoke heating reactor

    International Nuclear Information System (INIS)

    Tseng, C.M.; Lepp, R.M.

    1982-04-01

    A 2 MWt SLOWPOKE reactor, intended for commercial space heating, is being developed at the Chalk River Nuclear Laboratories. A small-signal dynamic simulation of this reactor, without closed-loop control, was developed. Basic equations were used to describe the physical phenomena in each kf the eight reactor subsystems. These equations were then linearized about the normal operation conditions and rearranged in a dimensionless form for implementation. The overall simulation is non-linear. Slow transient responses (minutes to days) of the simulation to both reactivity and temperature perturbations were measured at full power. In all cases the system reached a new steady state in times varying from 12 h to 250 h. These results illustrate the benefits of the inherent negative reactivity feedback of this reactor concept. The addition of closed-loop control using core outlet temperature as the controlled variable to move a beryllium reflector is also examined

  14. Distribution of energy of impulses of the modernized IBR-2 REACTOR

    International Nuclear Information System (INIS)

    Tayibov, L.A; Mehtiyeva, R.N.; )

    2011-01-01

    Full text: For the modernized IBR-2 reactor there are two main reasons causing fluctuations of energy of impulses [1,3] on low power of stochastic fluctuations, on the nominal - giving rise to fluctuations of external reactance. The fluctuations of pulse energy is quite significant (20%). They affect the dynamics of the reactor, the process of regulation, starting, as well as the work of the experimental apparatus, etc. It is clear that research of fluctuation of energy of impulses has special value for the IBR-2 type reactor. Sufficient information about the statistical properties of the reactor noise gives the density distribution of the energy pulse power. We used the usual procedure of statistical analysis of time series. Calculated pulse energy of density and the parameters of this distribution.

  15. Os custos eleitorais do Bolsa Família: reavaliando seu impacto sobre a eleição presidencial de 2006

    Directory of Open Access Journals (Sweden)

    Diego Sanches Corrêa

    2015-12-01

    Full Text Available O padrão geográfico da flutuação da votação de Lula entre as eleições presidenciais de 2002 e 2006 é um dos mais intrigantes fenômenos políticos da história recente brasileira. Diversos estudos mostram que o programa Bolsa Família aumentou consideravelmente o apoio a Lula entre os pobres, tendo um papel determinante nos resultados da eleição de 2006. Neste artigo, eu demonstro com base em um banco de dados municipais e técnicas de econometria espacial que seu desempenho eleitoral também se associa negativamente à proporção de ricos. Meu argumento é de que o programa explica ambos os efeitos: os pobres responderam às melhorias em suas condições materiais de vida e os ricos aos custos de oportunidade de investimentos públicos que não lhes beneficiaram diretamente.

  16. Impacto sobre el estado de salud de los programas de transferencia condicionada de renta: el Programa Bolsa Familia de Brasil

    Directory of Open Access Journals (Sweden)

    Berta Rivera Castiñeira

    2009-01-01

    Full Text Available Las transferencias condicionadas de renta se están consolidando como instrumento estándar para la reducción de la pobreza. El Programa Bolsa Familia implementado en Brasil es el de mayor envergadura de este tipo de programa en el mundo. La evaluación de su impacto ofrece algunas indicaciones extrapolables a otros países. En este artículo se pone en evidencia la falta de resultados de este programa en términos de estado de salud y de modificación de conductas no saludables. la existencia de barreras por el lado de la oferta aparece como la limitación más importante para la consecución de mejores resultados en este ámbito. Sin embargo, el impacto positivo del programa sobre la educación y la reducción de la pobreza permite predecir mejoras en el estado de salud de la población a largo plazo.

  17. [Bolsa-Família Program: diet quality of adult population in Curitiba, Paraná].

    Science.gov (United States)

    Lima, Flávia Emília Leite de; Fisberg, Regina Mara; Uchimura, Kátia Yumi; Picheth, Telma

    2013-03-01

    This study evaluated the quality of diet of the population receiving the Bolsa Familia Program in Curitiba, state of Parana, Brazil. It was a population-based cross-sectional study, conducted from July 2006 to July 2007. 747 beneficiaries were interviewed from 19 years of age, of both genders. A 24 hour-recall was implemented in order to assess the quality of the diet and the Healthy Eating Index (HEI) was used as a parameter for the classification of the group in consumption levels. Descriptive statistics were used to describe the diet quality of the studied population. Wald test and ANOVA test were performed to compare the means of the index according to the socio-economic variables, considering a significance level of 5%. The sample comprised 91.4% of women and 8.6% of men. The average age of the population was 36.4 ± 13.3 years, with 75% having completed elementary school. The mean HEI was 51 points, which features a diet that needs improvement. The population has a monotonous diet with an adequate intake of legumes, but low for fruits, vegetables and dairy products. Comparing the categories of diet quality of individuals, all components, except sodium, showed statistically different median score (p < 0.01). Studies that evaluate the quality of the diet are essential to support the implementation of nutrition education programs targeted to the core of the problem in the populations studied.

  18. Cartografias da cópia: estudo sobre o consumo subalterno de bolsas de luxo piratas

    Directory of Open Access Journals (Sweden)

    Carla Gavilan Carvalho

    2012-06-01

    Full Text Available Normal 0 21 false false false PT-BR X-NONE X-NONE /* Style Definitions */ table.MsoNormalTable {mso-style-name:"Tabela normal"; mso-tstyle-rowband-size:0; mso-tstyle-colband-size:0; mso-style-noshow:yes; mso-style-priority:99; mso-style-qformat:yes; mso-style-parent:""; mso-padding-alt:0cm 5.4pt 0cm 5.4pt; mso-para-margin-top:0cm; mso-para-margin-right:0cm; mso-para-margin-bottom:10.0pt; mso-para-margin-left:0cm; line-height:115%; mso-pagination:widow-orphan; font-size:11.0pt; font-family:"Calibri","sans-serif"; mso-ascii-font-family:Calibri; mso-ascii-theme-font:minor-latin; mso-fareast-font-family:"Times New Roman"; mso-fareast-theme-font:minor-fareast; mso-hansi-font-family:Calibri; mso-hansi-theme-font:minor-latin;} Este artigo pretende refletir sobre o consumo de bolsas de marcas luxuosas como prática social com implicações culturais e suas relações produzidas, a partir da perspectiva de que o consumo é capaz de definir modos de ser, trabalhar e atuar enquanto cidadão. Avalia também como tal prática tem resignificado o consumo tradicional, assim como a definição de luxo na sociedade contemporânea.

  19. Equipment for thermal neutron flux measurements in reactor R2

    Energy Technology Data Exchange (ETDEWEB)

    Johansson, E; Nilsson, T; Claeson, S

    1960-04-15

    For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

  20. A conceptual design of LIB fusion reactor: UTLIF(2)

    International Nuclear Information System (INIS)

    Madarame, Haruki; Kondo, Shunsuke; Iwata, Shuichi; Oka, Yoshiaki; Miya, Kenzo.

    1984-01-01

    UTLIF(2) is a conceptual design study on a light ion beam driven fusion reactor based on a concept of rod-bundle blanket. Survivability and maintainability of the first wall and the blanket are regarded as of major importance in the design. The blanket rod is composed of a thick tube which has enough stiffness, a thin wrapping wall which receives high heat flux, and liquid lithium which breeds tritium and removes generated heat. The rod can be pulled out from the outside of the reactor vessel, hence the replacement is very easy. Nuclear and thermal analysis have been made and the performance of the reactor has been shown to be satisfactory. (author)

  1. The BR2 materials testing reactor. Past, ongoing and under-study upgradings

    Energy Technology Data Exchange (ETDEWEB)

    Baugnet, J M; Roedt, Ch de; Gubel, P; Koonen, E [Centre d' Etude de I' Energie Nucleaire, Studiecentrum voor Kernenergie, C.E.N./S.C.K., Mol (Belgium)

    1990-05-01

    The BR2 reactor (Mol, Belgium) is a high-flux materials testing reactor. The fuel is 93% {sup 235}U enriched uranium. The nominal power ranges from 60 to 100 MW. The main features of the design are the following: 1) maximum neutron flux, thermal: 1.2 x 10{sup 15} n/cm{sup 2} s; fast (E > 0.1 MeV) : 8.4 x 10{sup 14} n /cm{sup 2} s; 2) great flexibility of utilization: the core configuration and operation mode can be adapted to the experimental loading; 3) neutron spectrum tailoring; 4) availability of five 200 mm diameter channels besides the standard channels (84 mm diameter); 5) access to the top and bottom covers of the reactor authorizing the irradiation of loops. The reactor is used to study the behaviour of fuel elements and structural materials intended for future nuclear power stations of several types (fission and fusion). Irradiations are carried out in connection with performance tests up to very high burn-up or neutron fluence as well as for safety experiments, power cycling experiments, and generally speaking, tests under off-normal conditions. Irradiations for nuclear transmutation (production of high specific activity radio-isotopes and transplutonium elements), neutron-radiography, use of beam tubes for physics studies, and gamma irradiations are also carried out. The BR2 is used in support of Belgian programs, at the request of utilities, industry and universities and in the framework of international agreements. The paper reviews the past and ongoing upgrading and enhancement of reactor capabilities as well as those under study or consideration, namely with regard to: reactor equipment, fuel elements, irradiation facilities, reactor operation conditions and long-term strategy. (author)

  2. TiO2 Solar Photocatalytic Reactor Systems: Selection of Reactor Design for Scale-up and Commercialization—Analytical Review

    Directory of Open Access Journals (Sweden)

    Yasmine Abdel-Maksoud

    2016-09-01

    Full Text Available For the last four decades, viability of photocatalytic degradation of organic compounds in water streams has been demonstrated. Different configurations for solar TiO2 photocatalytic reactors have been used, however pilot and demonstration plants are still countable. Degradation efficiency reported as a function of treatment time does not answer the question: which of these reactor configurations is the most suitable for photocatalytic process and optimum for scale-up and commercialization? Degradation efficiency expressed as a function of the reactor throughput and ease of catalyst removal from treated effluent are used for comparing performance of different reactor configurations to select the optimum for scale-up. Comparison included parabolic trough, flat plate, double skin sheet, shallow ponds, shallow tanks, thin-film fixed-bed, thin film cascade, step, compound parabolic concentrators, fountain, slurry bubble column, pebble bed and packed bed reactors. Degradation efficiency as a function of system throughput is a powerful indicator for comparing the performance of photocatalytic reactors of different types and geometries, at different development scales. Shallow ponds, shallow tanks and fountain reactors have the potential of meeting all the process requirements and a relatively high throughput are suitable for developing into continuous industrial-scale treatment units given that an efficient immobilized or supported photocatalyst is used.

  3. Planned Scientific programs around the Triga Mark 2 Reactor

    International Nuclear Information System (INIS)

    Majah, M Ibn.

    2007-01-01

    Full text: Nuclear techniques have been introduced to Morocco since the sixties. After the energy crisis of 1973, Morocco decides to create the National Center for Energy Sciences and Nuclear Techniques (CNESTEN) under the supervision of the Ministry of high Education and Research, with a research commercial and support vocation. CNESTEN is in charge of promoting nuclear application, to act as technical support for the authorities and to prepare the technological basis for nuclear power option. In 1998, CNESTEN started the construction of Nuclear Research Centre. The on going activities cover many sectors : earth and environmental sciences, high energy physics, safety and security, waste management. In 2001, CNESTEN started the construction of a 2MW TRiga Mark 2 Reactor, with the possibility to increase the power to 3 MW. The construction was achieved in January 2007. The operation of the reactor is expected for April 2007. The program of the utilization of the reactor was established with th contribution of the university and with the assistance of IAEA. Some of the experimental set-up installed around the reactor have been designed. CNESTEN has developed cooperation with Nuclear research centres from other countries and is receiving visitors and trainees mainly through the IAEA [fr

  4. Perfil dos pesquisadores com bolsa de produtividade em pesquisa do CNPq da área de saúde coletiva

    Directory of Open Access Journals (Sweden)

    Rita Barradas Barata

    2003-12-01

    Full Text Available O artigo analisa, com base nas informações do currículo Lattes, o perfil dos pesquisadores com bolsa de produtividade em pesquisa do CNPq na área de Saúde Coletiva. A análise levou em conta a formação graduada e pós-graduada, área de atuação, produção e divulgação científica. As comparações são feitas entre as classes de pesquisadores e com dados do diretório de grupos de pesquisa. A maioria dos pesquisadores (70% são formados em Ciências da Saúde, principalmente em Medicina, ou em Ciências Humanas (18%, principalmente Sociologia. Sessenta por cento fizeram mestrado e doutorado em Saúde Coletiva, mas há entre 20 e 30% de pesquisadores, dependendo da classe, sem formação específica na área. A maioria atua em Epidemiologia. A produção científica, expressa em produtos bibliográficos, varia de 10,56 produtos/ano de obtenção do doutorado para os pesquisadores 2C a 6,60 produtos/ano para os pesquisadores 1A. Para artigos completos publicados em periódicos os valores são 3,56 e 2,87, respectivamente. A produção é divulgada principalmente em periódicos A internacional e, A e B nacional. Os periódicos que concentram a publicação são Cadernos de Saúde Pública e Revista de Saúde Pública.

  5. Power noise spectrum classification in the problem of the IBR-2 reactor

    International Nuclear Information System (INIS)

    Bargel, M.; Kitowski, J.; Pepelyshev, Yu.N.

    1988-01-01

    The classification spectrum results of random fluctuations in the IBR-2 energy pulse are presented. The work is performed for the application of the obtained results to the reactor diagnostics and the study of its noise uncontrolled states. For classification of the spectra the method of pattern recognition based upon the ISODATA heuristic algorithm is used. It is shown that a set of noise uncontrolled reactor states, registered during the reactor operation period at power of 0.4-2 MVt with the first variant of moving reflector (1983-1986) is formed into 4(5) most typical states. Each of the states corresponds to the general conditions of the reactor core cooling and provides the normal work of the moving reflector. However, these states differ in coolant flow, power level and peculiarities of the moving reflector rotation regime. One type of anomal power noise, connected with some disorder in the moving reflctor work, is isolated. This work also presents the possibility of control over the state of moving reflectors according to the change in the amplitude of power oscillations at some frequences. The reactor noise classification results can be used as the data bank for the IBR-2 reactor diagnostic system

  6. Research reactor FR2 - 20 years chemical and radiochemical measurements

    International Nuclear Information System (INIS)

    Feuerstein, H.; Graebner, H.; Oschinski, J.; Hoffmann, W.; Beyer, J.

    1986-09-01

    The FR2 has been a D 2 O cooled and moderated research reactor with a thermal output of 44 MW. It was in operation from 1961 to 1981. Because of the operating conditions of the reactor, only a small number of routine measurements were performed. For these however special techniques had to be developed. During the 20 years of operation a number of special events occured or have been observed, sometimes with very amazing results, e.g. the 'aceton effect'. This report describes the chemical and radiochemical conditions of the reactor systems, as well as the results of the surveilance work. Not described are measurements for the many experiments. The last chapter gives in a short form a description of the most unusual events and observations. (orig.) [de

  7. Venting krypton-85 from the Three Mile Island Unit 2 reactor building

    International Nuclear Information System (INIS)

    Burton, H.M.

    1981-01-01

    To permit the less restricted access to the reactor building necessary to maintain instrumentation and equipment, and to proceed towad the total decontamination of the facility, General Public Utilities, operators of the facility referred to hereafter as GPU, asked the United States Nuclear Regulatory Commission, or NRC, for permission to remove the 85 Kr from the reactor building by venting it to the environment. GPU supported their request with the Safety Analysis and Environmental Assessment Report on the proposed reactor building venting plan. On June 12, 1980, after seven months of licensing deliberations and numerous public hearings, the NRC granted GPU's request. The actual venting took place between June 28 and July 11, 1980. This report presents an overview of the detailed effort involved in the TMI-2 reactor building venting program. The findings reported here are condensed from a published report entitled TMI-2 Reactor Building Purge--Kr-85 Venting

  8. Research on economics and CO2 emission of magnetic and inertial fusion reactors

    International Nuclear Information System (INIS)

    Mori, Kenjiro; Yamazaki, Kozo; Oishi, Tetsutarou; Arimoto, Hideki; Shoji, Tatsuo

    2011-01-01

    An economical and environment-friendly fusion reactor system is needed for the realization of attractive power plants. Comparative system studies have been done for magnetic fusion energy (MFE) reactors, and been extended to include inertial fusion energy (IFE) reactors by Physics Engineering Cost (PEC) system code. In this study, we have evaluated both tokamak reactor (TR) and IFE reactor (IR). We clarify new scaling formulas for cost of electricity (COE) and CO 2 emission rate with respect to key design parameters. By the scaling formulas, it is clarified that the plant availability and operation year dependences are especially dominant for COE. On the other hand, the parameter dependences of CO 2 emission rate is rather weak than that of COE. This is because CO 2 emission percentage from manufacturing the fusion island is lower than COE percentage from that. Furthermore, the parameters dependences for IR are rather weak than those for TR. Because the CO 2 emission rate from manufacturing the laser system to be exchanged is very large in comparison with CO 2 emission rate from TR blanket exchanges. (author)

  9. Development of a TiO2-coated optical fiber reactor for water decontamination

    International Nuclear Information System (INIS)

    Danion, A.

    2004-09-01

    The objective of this study was to built and to study a photo-reactor composed by TiO 2 -coated optical fibers for water decontamination. The physico-chemical characteristics and the optical properties of the TiO 2 coating were first studied. Then, the influences of different parameters as the coating thickness, the coating length and the coating volume were investigated both on the light transmission in the TiO 2 - coated fiber and on the photo-catalytic activity of the fiber for a model compound (malic acid). The photo-catalytic degradation of malic acid was optimized using the experimental design methodology allowing to build a multi-fiber reactor comprising 57 optical fibers. The photo-degradation of malic acid was conducted in the multi-fiber reactor and it was demonstrated that the multi-fiber reactor was more efficient than the single-fiber reactor at the same fibers density. Finally, the multi-fiber reactor was applied to the photo-degradation of a fungicide, called fenamidone, and a degradation pathway was proposed. (author)

  10. TMI-2 reactor-vessel head removal and damaged-core-removal planning

    International Nuclear Information System (INIS)

    Logan, J.A.; Hultman, C.W.; Lewis, T.J.

    1982-01-01

    A major milestone in the cleanup and recovery effort at TMI-2 will be the removal of the reactor vessel closure head, planum, and damaged core fuel material. The data collected during these operations will provide the nuclear power industry with valuable information on the effects of high-temperature-dissociated coolant on fuel cladding, fuel materials, fuel support structural materials, neutron absorber material, and other materials used in reactor structural support components and drive mechanisms. In addition, examination of these materials will also be used to determine accident time-temperature histories in various regions of the core. Procedures for removing the reactor vessel head and reactor core are presented

  11. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  12. Práctica preprofesional de enfermería familiar y calidad de vida en familias del Barrio Tola Chica, 2014

    Directory of Open Access Journals (Sweden)

    Christian Fernando Juna Juca

    2017-03-01

    Full Text Available Introducción: La Calidad de Vida Relacionada a la Salud es la percepción subjetiva, influenciada por el estado de salud actual, de la capacidad para realizar actividades importantes para el individuo. Objetivo: Determinar la influencia de las prácticas preprofesionales de Enfermería Familiar de la Pontificia Universidad Católica del Ecuador en la percepción de la Calidad de Vida Relacionada a Salud de las familias del barrio Tola Chica de Quito. Métodos: Se realizó un estudio cuasi-experimental con diseño de grupo control sin selección aleatoria, se aplicó la encuesta SF-12v2 para evaluar la Calidad de Vida Relacionada a Salud, se administró una encuesta sociodemográfica, una de estratificación socioeconómica y una lista de chequeo de intervenciones de enfermería. Fueron utilizados los softwares Quality Metric Health Outcomes, SPSS 20.0 y JMP® 9.0.1 para el cálculo y asociación de variables estudiadas. Resultados: Predominó el tipo de familia nuclear 68,9% y el estrato socioeconómico C-(64,4%. La actividad que realizaron los estudiantes con mayor frecuencia, educación sanitaria (95,6%; incremento del afrontamiento (91,9% y potenciación de la socialización (91,7%. Se influyó en la Calidad de Vida Relacionada a Salud, en el componente físico (F = 26,19 GL = 44 p < 0,001 y en el componente mental (F = 54,49 GL = 44 p < 0,001.  Conclusiones: La planificación de los objetivos de práctica propuestos al inicio del periodo académico, en función de los componentes de la Calidad de Vida Relacionada a Salud, permitió un incremento de los dominios físico y mental en general.

  13. Fissile fuel production and usage of thermal reactor waste fueled with UO2 by means of hybrid reactor system

    International Nuclear Information System (INIS)

    Ipek, O.

    1997-01-01

    The use of Fast Breeder Reactors to produce fissile fuel from nuclear waste and the operation of these reactors with a new neutron source are becoming today' topic. In the thermonuclear reactors, it is possible to use 2.45-14.1 MeV - neutrons which can be obtained by D-T, D-D Semicatalyzed (D-D) and other fusion reactions. To be able to do these, Hybrid Reactor System, which still has experimental and theoretical studies, have to be taken into consideration.In this study, neutronic analysis of hybrid blanket with grafit reflector, is performed. D-T driven fusion reaction is surrounded by UO 2 fuel layer and the production of ''2''3''9Pu fissile fuel from waste ''2''3''8U is analyzed. It is also compared to the other possible fusion reactions. The results show that 815.8 kg/year ''2''3''8Pu with D-T reaction and 1431.6 kg/year ''2''3''8Pu with semicatalyzed (D-D) reaction can be produced for 1000 MW fusion power. This means production of 2.8/ year and 4.94/ year LWR respectively. In addition, 1000 MW fusion flower is is multiplicated to 3415 MW and 4274 MW for D-T and semicatalyzed (D-D) reactions respectively. The system works subcritical and these values are 0.4115 and 0.312 in order. The calculations, ANISN-ORNL code, S 16 -P 3 approach and DLC36 data library are used

  14. A theoretical analysis of methanol synthesis from CO2 and H2 in a ceramic membrane reactor

    NARCIS (Netherlands)

    Gallucci, F.; Basile, A.

    2007-01-01

    In this theoretical work the CO2 conversion into methanol in both a traditional reactor (TR) and a membrane reactor (MR) is considered. The purpose of this study was to investigate the possibility of increasing CO2 conversion into methanol with respect to a TR. A zeolite MR, able to combine

  15. Cronos 2: a neutronic simulation software for reactor core calculations

    International Nuclear Information System (INIS)

    Lautard, J.J.; Magnaud, C.; Moreau, F.; Baudron, A.M.

    1999-01-01

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  16. Set of rules SOR 2 reactor site criteria

    International Nuclear Information System (INIS)

    1976-06-01

    The purpose of this set of rules is to describe criteria which guide the Director in his evaluation of the suitability of proposed sites for stationary power and testing reactors subject to SOR 2. (B.G.)

  17. Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    Liu Guisheng

    1995-01-01

    How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testings of CENDL-2 and ENDF/B-6. (4 tabs., 2 figs.)

  18. Groundwater Monitoring Plan for the Reactor Technology Complex Operable Unit 2-13

    International Nuclear Information System (INIS)

    Richard P. Wells

    2007-01-01

    This Groundwater Monitoring Plan describes the objectives, activities, and assessments that will be performed to support the on-going groundwater monitoring requirements at the Reactor Technology Complex, formerly the Test Reactor Area (TRA). The requirements for groundwater monitoring were stipulated in the Final Record of Decision for Test Reactor Area, Operable Unit 2-13, signed in December 1997. The monitoring requirements were modified by the First Five-Year Review Report for the Test Reactor Area, Operable Unit 2-13, at the Idaho National Engineering and Environmental Laboratory to focus on those contaminants of concern that warrant continued surveillance, including chromium, tritium, strontium-90, and cobalt-60. Based upon recommendations provided in the Annual Groundwater Monitoring Status Report for 2006, the groundwater monitoring frequency was reduced to annually from twice a year

  19. Diseño de un seguro para reducir el riesgo ante las variaciones de los precios de las acciones en la bolsa de valores de lima y desarrollar el Mercado de capitales

    OpenAIRE

    Bacigalupo Pozo, Juan Alberto

    2016-01-01

    La presente tesis, tuvo por objetivo diseñar un seguro para reducir el riesgo de los inversionistas por las variaciones de precios en el mercado de acciones en la Bolsa de Valores de Lima y a partir de ello, desarrollar el mercado de capitales. Esto se hizo posible a través de la aplicación de la hoja de encuesta a los inversionistas seleccionados (utilizando el criterio de exclusión); asimismo, se desarrolló un modelo econométrico y se ejecutó una simulación aleatoria para pro...

  20. Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    Liu Guisheng

    1995-11-01

    How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testing of CENDL-2 and ENDF/B-6. (author). 8 refs, 2 figs, 4 tabs

  1. Hamster (Mesocricetus auratus cheek pouch as an experimental model to investigate human skin and keloid heterologous graft Bolsa jugal no hamster (Mesocricetus auratus como modelo experimental de investigação de enxertos heterólogos de pele humana e quelóide

    Directory of Open Access Journals (Sweden)

    Bernardo Hochman

    2004-12-01

    Full Text Available To describe the integration process of grafts of total human skin and keloid in hamster (Mesocricetus auratus cheek pouch, whose sub-epithelium is naturally an "Immunologically Privileged Site". Fragments of human normal skin and keloid from the breast region of mulatto female patients were transplanted into the cheek pouch subepithelium in situ. Surgical procedure and grafted pouches for microscopic exam at several time points of the transplantation were standardized. The integration of grafted fragments of human skin and keloid was seen in late periods (84 days since the microscopic assessment showed the presence of blood vases within the conjunctive tissue of grafted fragments. It was also possible to see among the grafted fragments the epithelium, the appearing of early cellular infiltrated, epithelial secretion of keratin, the presence of melanocytes, and delayed changes on the aspect of collagen fibers of conjunctive tissue. Pooled results allow to define hamster cheek pouch sub-epithelium as an experimental model to investigating heterologous graft physiology of human total skin and keloid with epithelium.Descrever a integração dos enxertos de pele total humana e quelóide na bolsa jugal do hamster (Mesocricetus auratus, cujo subepitélio é, naturalmente, um "Local de Privilégio Imunológico". Foram transplantados fragmentos, de pele humana normal e de quelóide, obtidos da região mamária de pacientes pardas, no subepitélio da bolsa jugal in situ. O procedimento operatório, e de preparo das bolsas enxertadas para exame microscópico em vários períodos de transplante, foi padronizado. Verificou-se a integração dos fragmentos enxertados de pele humana e de quelóide em períodos tardios (84 dias, uma vez que a avaliação microscópica revelou a presença de vasos sangüíneos no tecido conjuntivo dos fragmentos enxertados. Foi também possível observar, nos fragmentos enxertados, o epitélio, o aparecimento de infiltrado

  2. An experimental investigation of fission product release in SLOWPOKE-2 reactors - Data report

    International Nuclear Information System (INIS)

    Harnden, A.M.C.

    1995-09-01

    The results of an investigation into the release of fission products from SLOWPOKE-2 reactors fuelled with a highly-enriched uranium alloy core are detailed in Volume 1. This data report (Volume 2) contains plots of the activity concentrations of the fission products observed in the reactor container at the University of Toronto, Ecole Polytechnique and the Kanata Isotope Production Facility. Release rates from the reactor container water to the gas headspace are also included. (author)

  3. Safety assessments relating to the use of new fuels in research reactors: application to the case of FRM 2 reactor fuel

    International Nuclear Information System (INIS)

    Abou Yehia, H.; Bars, G.; Tran Dai

    2001-01-01

    After giving a brief reminder of the procedure applied in France for the licensing of the use of a new fuel type or design in a research reactor, we outline the main safety aspects associated with such a modification. Finally, by way of an example, we focus on the safety assessment relating to the IRIS irradiation device used in SILOE reactor, in particular for the qualification of the fuel dedicated to FRM II reactor of the Technical University of Munich. This qualification was carried out on a U 3 Si 2 fuel plate enriched to about 90 % in weight of 235 U and containing 1.5 g of uranium per cm 3 . The evaluation performed by the IPSN for GRS did not call into question the choice of U 3 Si 2 fuel plates for the FRM-II reactor. (authors)

  4. Thermal neutron flux distribution in ET-RR-2 reactor thermal column

    Directory of Open Access Journals (Sweden)

    Imam Mahmoud M.

    2002-01-01

    Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

  5. Apollo-L2, an advanced fuel tokamak reactor utilizing direct conversion

    International Nuclear Information System (INIS)

    Emmert, G.A.; Kulcinski, G.L.; Blanchard, J.P.; El-Guebaly, L.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.E.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Witt, R.J.

    1989-01-01

    A scoping study of a tokamak reactor fueled by a D- 3 He plasma is presented. The Apollo D- 3 He tokamak capitalizes on recent advances in high field magnets (20 T) and utilizes rectennas to convert the synchrotron radiation directly to electricity. The low neutron wall loading (0.1 MW/m 2 ) permits a first wall lasting the life of the plant and enables the reactor to be classified as inherently safe. The cost of electricity is less than that from a similar power level DT reactor. 10 refs., 1 fig., 4 tabs

  6. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    Kouts, H.

    1986-01-01

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

  7. Rapid data acquisition from the safety system of the FRJ-2 reactor

    International Nuclear Information System (INIS)

    Inhoven, H.

    1980-06-01

    The central department for research reactors (ZFR) of the Juelich Nuclear Research Centre (KFA) is operating the reactors FRJ-1 (MERLIN) and FRJ-2 (DIDO) since 1962. In 1976, a Siemens 330 computer has been put into operation especially for the processing of data from the DIDO reactor, followed by another computer of the same type for the purpose of processing data from the ZFR department in general. The present report is a result of the work investigating 'Data acquisition and data processing in the FRJ-2' and primarily discusses the complex of 'fast analog and binary signals'. The activities in this field of work have been and still are mainly concerned with general problems encountered in adapting a currently 14-year-old reactor system to a digital computer, namely problems such as data decoupling in the safety system of the reactor, data acquisition using the CAMAC system, data transfer via an 'extended branch', data acquisition software as core-resident programs, temporary storage as common data, interpreting software as peripheral - storage - resident programs. (orig./WB) [de

  8. LA ADOLESCENCIA ANTE LA VIOLENCIA DE GÉNERO 2.0: CONCEPCIONES, CONDUCTAS Y EXPERIENCIAS

    Directory of Open Access Journals (Sweden)

    Trinidad Donoso Vázquez

    2018-01-01

    Full Text Available El objetivo del estudio es presentar un instrumento de medida sobre las vio- lencias de género 2.0 y analizar las percepciones y experiencias de adolescen- tes, así como las respuestas ante tales violencias. Se ha realizado un estudio por encuesta a 3.043 adolescentes de Cataluña, Aragón, Galicia, Andalucía, Islas Baleares y Canarias. Entre los resultados encontrados destaca que los y las adolescentes creen que hay más violencia online que of ̄ine; y perciben las acciones más directas y evidentes ligadas a la violencia sexual, pero en cambio las conductas que menos se perciben como violentas son aquellas en las que la mujer es tratada como objeto sexual, y especialmente no se perci- ben como violentas las conductas de control que se ejercen sobre la pareja a través de los entornos virtuales. La tendencia a la agresión es masculina, pero las chicas muestran más conductas agresoras en violencias relacionados con los mitos del amor romántico. Las respuestas de las y los adolescentes son pa- sivas ante las violencias de género 2.0, aunque las chicas superan a los chicos en las respuestas activas frente a la violencia. Se concluye con la necesidad de realizar intervenciones educativas para preparar a las y los adolescentes ante las violencias de género 2.0. Intervenciones que deberán estar orientadas a: deconstruir los mitos del amor romántico que sustentan falsas ideas sobre la con®anza en la pareja, a concienciar a las chicas sobre su mayor vulnera- bilidad en las redes sociales, a explicar a los y las adolescentes los riesgos de todas las redes sociales, y a alentar a los jóvenes a denunciar las acciones de violencia e implicar a la familia y la escuela en estos asuntos.

  9. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  10. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

  11. Estimation of power feedback parameters of pulse reactor IBR-2M on transients

    International Nuclear Information System (INIS)

    Pepyolyshev, Yu.N.; Popov, A.K.

    2013-01-01

    Parameters of the IBR-2M reactor power feedback (PFB) on a model of the reactor dynamics by mathematical treatment of two registered transients are estimated. Frequency characteristics and the pulse transient characteristics corresponding to these PFB parameters are calculated. PFB parameters received thus can be considered as their express tentative estimation as real measurements in this case occupy no more than 30 minutes. Total PFB is negative at 1 and 2 MW. At the received estimations of PFB parameters in a self-regulation mode it is possible to consider the stability margins of the IBR-2M reactor satisfactory

  12. Independent CO2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Task 2.50.05

    International Nuclear Information System (INIS)

    Stojic, M.; Pavicevic, M.

    1964-01-01

    This report contains the following volumes V and VI of the Project 'Independent CO 2 loop for cooling the samples irradiated in RA reactor vertical experimental channels': Design project of the dosimetry control system in the independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, and Safety report for the Independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels [sr

  13. PSA Level 2 activities for RBMK reactors

    International Nuclear Information System (INIS)

    Gubler, R.

    1998-01-01

    Probabilistic safety analyses (PSAs) of the boiling water graphite moderated pressure tube reactors (RBMKs) have been developed only recently and they are limited to Level 1. Activities at the IAEA were first motivated because of the difficulties to characterize core damage for RBMK reactors. Core damage probability is used in documents of the IAEA as a convenient single valued measure, for example for probabilistic safety criteria. The limited number of PSAs that have been completed for the RBMK reactors have shown that several special features of these channel type reactors necessitate revisiting of the characterization of core damage for these reactors. Furthermore, it has become increasingly evident that detailed deterministic analysis of DBAs and beyond design basis accidents reveal considerable insights into RBMK response to various accident conditions. These analyses can also help in better characterizing the outstanding phenomenological uncertainties, improved EOPs and AM strategies, including potential risk-beneficial accident negative backfits. The deterministic efforts should be focused first on elucidating accident progression processes and phenomena, and second on finding, qualifying and implementing procedures to minimize the risk of severe accident states The IAEA PSA procedures were mainly developed in New of vessel type LWRs, and would therefore require extensions to make them directly applicable. to channel type reactors. (author) (author)

  14. Studsvik's R2 reactor - Review of activities

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, Mikael; Tomani, Hans; Graeslund, Christian; Rundquist, Hans; Skoeld, Kurt [Studsvik Nuclear AB, Nykoeping (Sweden)

    1993-07-01

    A general description of the R2 reactor, its associated facilities and its history is given. The facilities and range of work are described for the following types of activities: fuel testing, materials testing, neutron transmutation doping of silicon, activation analysis, radioisotope production and basic research including thermal neutron scattering, nuclear chemistry and neutron capture radiography. (author)

  15. Decommissioning of reactor facilities (2). Required technology

    International Nuclear Information System (INIS)

    Yanagihara, Satoshi

    2014-01-01

    Decommissioning of reactor facilities was planned to perform progressive dismantling, decontamination and radioactive waste disposal with combination of required technology in a safe and economic way. This article outlined required technology for decommissioning as follows: (1) evaluation of kinds and amounts of residual radioactivity of reactor facilities with calculation and measurement, (2) decontamination technology of metal components and concrete structures so as to reduce worker's exposure and production of radioactive wastes during dismantling, (3) dismantling technology of metal components and concrete structures such as plasma arc cutting, band saw cutting and controlled demolition with mostly remote control operation, (3) radioactive waste disposal for volume reduction and reuse, and (4) project management of decommissioning for safe and rational work to secure reduction of worker's exposure and prevent the spreading of contamination. (T. Tanaka)

  16. Reproduction of the PSBR reactor with Exterminator-2; Reproduccion del reactor PSBR con exterminador-2

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1983-08-15

    To reproduce the reactor PSBR reported in (1), with the available version of the Exterminator-II in the ININ, they took the dimensions, composition specifications, effective sections of the different compositions (excepting those of the central thimble and of the moderator), the K{sub eff} and the factors of power (FP) for the different burners. Based on the comparison of the K{sub eff} and of the FP obtained with those reported the precision it is determined before in the reproduction of the reactor mentioned. (Author)

  17. Production of Sn-117m in the BR2 high-flux reactor.

    Science.gov (United States)

    Ponsard, B; Srivastava, S C; Mausner, L F; Russ Knapp, F F; Garland, M A; Mirzadeh, S

    2009-01-01

    The BR2 reactor is a 100MW(th) high-flux 'materials testing reactor', which produces a wide range of radioisotopes for various applications in nuclear medicine and industry. Tin-117m ((117m)Sn), a promising radionuclide for therapeutic applications, and its production have been validated in the BR2 reactor. In contrast to therapeutic beta emitters, (117m)Sn decays via isomeric transition with the emission of monoenergetic conversion electrons which are effective for metastatic bone pain palliation and radiosynovectomy with lesser damage to the bone marrow and the healthy tissues. Furthermore, the emitted gamma photons are ideal for imaging and dosimetry.

  18. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  19. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

  20. Comparative study between fluidized bed and fixed bed reactors in methane reforming with CO2 and O2 to produce syngas

    International Nuclear Information System (INIS)

    Jing Qiangshan; Lou Hui; Mo Liuye; Zheng Xiaoming

    2006-01-01

    Reforming of methane with carbon dioxide and oxygen was investigated over Ni/MgO-SiO 2 catalysts using fixed bed and fluidized bed reactors. The conversions of CH 4 and CO 2 in a fluidized bed reactor were close to thermodynamic equilibrium. The activity and stability of the catalyst in the fixed bed reactor were lower than that in the fluidized bed reactor due to carbon deposition and nickel sintering. TGA and TEM techniques were used to characterize the spent catalysts. The results showed that a lot of whisker carbon was found on the catalyst in the rear of the fixed bed reactor, and no deposited carbon was observed on the catalysts in the fluidized bed reactor after reaction. It is suggested that this phenomenon is related to a permanent circulation of catalyst particles between the oxygen rich and oxygen free zones. That is, fluidization of the catalysts in the fluidized bed reactor favors inhibiting deposited carbon and thermal uniformity in the reactor

  1. Benchmark testing of Canadol-2.1 for heavy water reactor

    International Nuclear Information System (INIS)

    Liu Ping

    1999-01-01

    The new version evaluated nuclear data library of ENDF-B 6.5 has been released recently. In order to compare the quality of evaluated nuclear data CENDL-2.1 with ENDF-B 6.5, it is necessary to do benchmarks testing for them. In this work, CENDL-2.1 and ENDF-B 6.5 were used to generated the WIMS-69 group library respectively, and benchmarks testing was done for the heavy water reactor, using WIMS5A code. It is obvious that data files of CENDL-2.1 is better than that of old WIMS library for the heavy water reactors calculations, and is in good agreement with those of ENDF-B 6.5

  2. Oxygen suppression in boiling water reactors. Phase 2. Annual report 1981, December 2, 1980-December 31, 1981

    International Nuclear Information System (INIS)

    Burley, E.L.

    1982-07-01

    A hydrogen addition test will be performed in the Dresden-2 reactor of Commonwealth Edison Company during 1982. Up to 2 ppM hydrogen will be added to and dissolved in the reactor feedwater to reverse the radiolysis reaction in the reactor core and suppress oxgen concentration in the primary coolant. At low oxygen levels the propensity of stressed and sensitized 304 stainless steel toward intergranular stress corrosion cracking is greatly reduced. The test will answer outstanding questions and uncertainties in the areas of water chemistry, equipment design and materials performance. Nine special sample facilities will be prepared in the primary coolant, main stream, feedwater/condensate, and offgas systems. Instrumentation will be available to measure hydrogen, oxygen, conductivity, pH, soluble and insoluble corrosion products, and electrochemical potentials. In addition, an autoclave in which confirming constant extension rate tests can be conducted in reactor water will be provided

  3. Characterization of fuel distributions in the Three-Mile Island Unit 2 (TMI-2) reactor system by neutron and gamma-ray dosimetry

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McNeece, J.P.; Kaiser, B.J.; McElroy, W.N.

    1984-04-01

    The resolution of technical issues generated by the accident at Three-Mile Island Unit 2 (TMI-2) will inevitably be of long range benefit. Determination of the fuel debris dispersal in the TMI-2 reactor system represents a major technical issue. In reactor recovery operations, such as for the safe handling and final disposal of TMI-2 waste, quantitative fuel assessments are being conducted throughout the reactor core and primary coolant system

  4. Programa Bolsa-Família: qualidade da dieta de população adulta do município de Curitiba, PR Bolsa-Família Program: Diet quality of adult population in Curitiba, Paraná

    Directory of Open Access Journals (Sweden)

    Flávia Emília Leite de Lima

    2013-03-01

    Full Text Available Este estudo avaliou a qualidade da dieta da população beneficiária do Programa Bolsa-Família, em Curitiba, PR. Estudo transversal, de base populacional, realizado no período de julho de 2006 a julho de 2007. Foram entrevistados 747 beneficiários, a partir dos 19 anos de idade, de ambos os sexos. Para avaliação da qualidade da dieta foi aplicado recordatório de 24 horas, e o Índice de Qualidade da Dieta (IQD foi utilizado como parâmetro para classificação do grupo em níveis de consumo. Estatística descritiva foi utilizada para descrever a qualidade da dieta da população. Para a comparação de médias do índice segundo as variáveis socioeconômicas foram realizados o teste t de Wald e a análise de variância ANOVA, considerando-se um nível de significância de 5%. A amostra foi constituída por 91,4% de mulheres e 8,6 % de homens. A média de idade da população foi de 36,4 ± 13,3 anos, com cerca de 75 % possuindo o ensino fundamental incompleto. A média do IQD foi de 51 pontos, o que caracteriza uma dieta que precisa de ajustes. A população possui uma dieta monótona, com um consumo adequado de leguminosas, porém baixo para frutas, verduras e produtos lácteos. Na comparação entre as categorias de qualidade da dieta dos indivíduos, todos os componentes, com exceção do sódio, apresentaram medianas de pontuação estatisticamente diferentes (p This study evaluated the quality of diet of the population receiving the Bolsa Familia Program in Curitiba, state of Parana, Brazil. It was a population-based cross-sectional study, conducted from July 2006 to July 2007. 747 beneficiaries were interviewed from 19 years of age, of both genders. A 24 hour-recall was implemented in order to assess the quality of the diet and the Healthy Eating Index (HEI was used as a parameter for the classification of the group in consumption levels. Descriptive statistics were used to describe the diet quality of the studied population. Wald

  5. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    International Nuclear Information System (INIS)

    Bissani, M; O'Kelly, D S

    2006-01-01

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to provide color-enhanced gemstones but is

  6. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Bissani, M; O' Kelly, D S

    2006-05-08

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to

  7. analysis and implementation of reactor protection system circuits - case study Egypt's 2 nd research reactor-

    International Nuclear Information System (INIS)

    Elnokity, O.E.M.

    2006-01-01

    this work presents a way to design and implement the trip unit of a reactor protection system (RPS) using a field programmable gate arrays (FPGA). instead of the traditional embedded microprocessor based interface design method, a proposed tailor made FPGA based circuit is built to substitute the trip unit (TU), which is used in Egypt's 2 nd research reactor ETRR-2. the existing embedded system is built around the STD32 field computer bus which is used in industrial and process control applications. it is modular, rugged, reliable, and easy-to-use and is able to support a large mix of I/O cards and to easily change its configuration in the future. therefore, the same bus is still used in the proposed design. the state machine of this bus is designed based around its timing diagrams and implemented in VHDL to interface the designed TU circuit

  8. Seguranca alimentar, renda e Programa Bolsa Familia: estudo de coorte em municipios do interior da Paraiba, Brasil, 2005-2011

    Directory of Open Access Journals (Sweden)

    Caroline Sousa Cabral

    2014-02-01

    Full Text Available Este trabalho tem por objetivo avaliar o impacto do Programa Bolsa Família na superação da Insegurança Alimentar. Realizou-se um estudo de coorte em 2005 e 2011, em amostra de famílias residentes em São José dos Ramos e Nova Floresta, Paraíba, Brasil. Em 2005 foram avaliados 609 domicílios e em 2011 foram encontradas e entrevistadas 406 famílias. Houve aumento da segurança alimentar/insegurança alimentar leve e melhoria nos indicadores socioeconômicos. Percebeu-se uma relação significativa entre a elevação da renda e a melhoria dos níveis de Insegurança Alimentar. O programa impacta positivamente no aumento da renda, propiciando melhorias dos níveis de segurança alimentar/insegurança alimentar leve. Percebeu-se que outras variáveis socioeconômicas podem estar contribuindo na melhoria deste perfil. Diante disso, no combate à insegurança alimentar e nutricional, são necessárias outras políticas e programas que ajam nos demais determinantes.

  9. General outline of the operation and utilization of the BR2 reactor

    International Nuclear Information System (INIS)

    Baugnet, J.M.; Leonard, F.; Gandolfo, J.M.; Lenders, H.

    1978-01-01

    The BR2 reactor is a high-flux material testing reactor of the thermal heterogeneous type. The fuel is 93% 235 U enriched uranium in the form of plates clad in aluminium. The moderator consists of beryllium and light water, the water being pressurized (12.5kg/cm 2 )and acting also as coolant. The pressure vessel is of aluminium, and is placed in a pool of demineralized water. One should stress the following main features of the design: the experimental channels are skew, the tube bundle presenting the form of a hyperboloid of revolution (see figure 1)-this gives easy access at the top and bottom reactor covers allowing complex instrumented devices, while maintaining a very high neutron flux at the core; great flexibilty of utilization, due to the fact that it is possible to adapt the core configuration to the experimental loading as the fissile charge can be centred on different experimental channels; although BR2 is a thermal reactor, it is possible to achieve neutron spectra very similar to those obtained in a fast reactor, either by the use of absorbing screens or by the use of fissile material within the experimental device; five 200mm diameter channels are available for loading large experimental irradiation devices, as in-pile sodium, gas or water loops. (author)

  10. Benchmark testing of CENDL-2 for U-fuel thermal reactors

    International Nuclear Information System (INIS)

    Zhang Baocheng; Liu Guisheng; Liu Ping

    1995-01-01

    Based on CENDL-2, NJOY-WIMS code system was used to generate 69-group constants, and do benchmark testing for TRX-1,2; BAPL-UO-2-1,2,3; ZEEP-1,2,3. All the results proved that CENDL-2 is reliable for thermal reactor calculations. (3 tabs.)

  11. Shadow corrosion testing in the INCA facility in the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Nystrand, A.C.; Lassing, A.

    1999-01-01

    Shadow corrosion is a phenomenon which occurs when zirconium alloys are in contact with or in proximity to other metallic objects in a boiling water reactor environment (BWR, RBMK, SGHWR etc.). An enhanced corrosion occurs on the zirconium alloy with the appearance of a 'shadow' of the metallic object. The magnitude of the shadow corrosion can be significant, and is potentially limiting for the lifetime of certain zirconium alloy components in BWRs and other reactors with a similar water chemistry. In order to evaluate the suitability of the In-Core Autoclave (INCA) in the Studsvik R2 materials testing reactor as an experimental facility for studying shadow corrosion, a demonstration test has been performed. A number of test specimens consisting of Zircaloy-2 tubing in contact with Inconel were exposed in an oxidising water chemistry. Some of the specimens were placed within the reactor core and some above the core. The conclusion of this experiment after post irradiation examination is that it is possible to use the INCA facility in the Studsvik R2 reactor to develop a significant level of shadow corrosion after only 800 hours of irradiation. (author)

  12. The 5th surveillance testing for Kori unit 2 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

  13. Digital, remote control system for a 2-MW research reactor

    International Nuclear Information System (INIS)

    Battle, R.E.; Corbett, G.K.

    1988-01-01

    A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs

  14. A nodal Grean's function method of reactor core fuel management code, NGCFM2D

    International Nuclear Information System (INIS)

    Li Dongsheng; Yao Dong.

    1987-01-01

    This paper presents the mathematical model and program structure of the nodal Green's function method of reactor core fuel management code, NGCFM2D. Computing results of some reactor cores by NGCFM2D are analysed and compared with other codes

  15. RHTF 2, a 1200 MWe high temperature reactor

    International Nuclear Information System (INIS)

    Brisbois, Jacques

    1978-01-01

    After having adapted to French conditions the 1160 MWe G.A.C. reactor, Commissariat a l'Energie Atomique and French Industry have decided to design an High Temperature Reactor 1200 MWe based on the G.A.C. technology and taking into account the point of view of Electricite de France and the experience of C.E.A. and industry on the gas cooled reactor technology. The main objective of this work is to produce a reactor design having a low technical risk, good operability, with an emphasis on the safety aspects easing the licensing problems

  16. Turkey's regulatory plans for high enriched to low enriched conversion of TR-2 reactor core

    International Nuclear Information System (INIS)

    Guelol Oezdere, Oya

    2003-01-01

    Turkey is a developing country and has three nuclear facilities two of which are research reactors and one pilot fuel production plant. One of the two research reactors is TR-2 which is located in Cekmece site in Istanbul. TR-2 Reactor's core is composed of both high enriched and low enriched fuel and from high enriched to low enriched core conversion project will take place in year 2005. This paper presents the plans for drafting regulations on the safety analysis report updates for high enriched to low enriched core conversion of TR-2 reactor, the present regulatory structure of Turkey and licensing activities of nuclear facilities. (author)

  17. Techno-economic assessment of membrane assisted fluidized bed reactors for pure H_2 production with CO_2 capture

    International Nuclear Information System (INIS)

    Spallina, V.; Pandolfo, D.; Battistella, A.; Romano, M.C.; Van Sint Annaland, M.; Gallucci, F.

    2016-01-01

    Highlights: • Membrane reactors improve the overall efficiency of H_2 production up to 20%. • Respect to conventional reforming, the H_2 yield increases from 12% to 20%. • The COH is reduced of at least 220% using membrane reactors. • FBMR capture 72% of CO_2 with a specific cost of 8 eur/tonn_C_O_2_. • MA-CLR can reach 90% of CO_2 avoided with same cost of FTR. - Abstract: This paper addresses the techno-economic assessment of two membrane-based technologies for H_2 production from natural gas, fully integrated with CO_2 capture. In the first configuration, a fluidized bed membrane reactor (FBMR) is integrated in the H_2 plant: the natural gas reacts with steam in the catalytic bed and H_2 is simultaneously separated using Pd-based membranes, and the heat of reaction is provided to the system by feeding air as reactive sweep gas in part of the membranes and by burning part of the permeated H_2 (in order to avoid CO_2 emissions for heat supply). In the second system, named membrane assisted chemical looping reforming (MA-CLR), natural gas is converted in the fuel rector by reaction with steam and an oxygen carrier (chemical looping reforming), and the produced H_2 permeates through the membranes. The oxygen carrier is re-oxidized in a separate air reactor with air, which also provides the heat required for the endothermic reactions in the fuel reactor. The plants are optimized by varying the operating conditions of the reactors such as temperature, pressures (both at feed and permeate side), steam-to-carbon ratio and the heat recovery configuration. The plant design is carried out using Aspen Simulation, while the novel reactor concepts have been designed and their performance have been studied with a dedicated phenomenological model in Matlab. Both configurations have been designed and compared with reference technologies for H_2 production based on conventional fired tubular reforming (FTR) with and without CO_2 capture. The results of the analysis show

  18. Reactor building integrity testing: A novel approach at Gentilly 2 - principles and methodology

    International Nuclear Information System (INIS)

    Collins, N.; Lafreniere, P.

    1991-01-01

    In 1987, Hydro-Quebec embarked on an ambitious development program to provide the Gentilly 2 nuclear power station with an effective, yet practical reactor building Integrity Test. The Gentilly 2 Integrity Test employs an innovative approach based on the reference volume concept. It is identified as the Temperature Compensation Method (TCM) System. This configuration has been demonstrated at both high and low test pressure and has achieved extraordinary precision in the leak rate measurement. The Gentilly 2 design allows the Integrity Test to be performed at a nominal 3 kPa(g) test pressure during an (11) hour period with the reactor at full power. The reactor building Pressure Test by comparison, is typically performed at high pressure 124 kPa(g)) in a 7 day window during an annual outage. The Integrity Test was developed with the goal of demonstrating containment availability. Specifically it was purported to detect a leak or hole in the 'bottled-up' reactor building greater in magnitude than an equivalent pipe of 25 mm diameter. However it is considered feasible that the high precision of the Gentilly 2 TCM System Integrity Test and a stable reactor building leak characteristic will constitute sufficient grounds for the reduction of the Pressure Test frequency. It is noted that only the TCM System has, to this date, allowed a relevant determination of the reactor building leak rate at a nominal test pressure of 3 kPa(g). Classical method tests at low pressure have lead to inconclusive results due to the high lack of precision

  19. Maximum credible accident analysis for TR-2 reactor conceptual design

    International Nuclear Information System (INIS)

    Manopulo, E.

    1981-01-01

    A new reactor, TR-2, of 5 MW, designed in cooperation with CEN/GRENOBLE is under construction in the open pool of TR-1 reactor of 1 MW set up by AMF atomics at the Cekmece Nuclear Research and Training Center. In this report the fission product inventory and doses released after the maximum credible accident have been studied. The diffusion of the gaseous fission products to the environment and the potential radiation risks to the population have been evaluated

  20. UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

    Energy Technology Data Exchange (ETDEWEB)

    Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra; Su’ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Aji, Indarta K. [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

    2016-03-11

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

  1. The Chernobyl reactor accident. Pt. 1 and 2

    International Nuclear Information System (INIS)

    1986-06-01

    The report first summarizes the available information on the various incidents of the whole accident scenario, and combines the information to present a first general outline and a basis for appraisal. The most significant incidents reported, namely power excursion, core meltdown, and fire, are discussed with a view to the reactor design and safety of reactors installed in the FRG. The main differences and advantages of German reactor designs are shown, as e.g.: Power excursions are mastered by inherent physical conditions; far better redundancy of engineered safety systems; enclosure of the complete reactor cooling system in a pressure-retaining steel containment; reactor buildings being made of reinforced concrete. The second part of the report deals with the radiological effects to be expected for our country. Data are given on the varying radiological exposure of the different regions. The fate and uptake of radioactivity in the human body are discussed. The conclusion drawn from the data presented is that the individual exposure due to the reactor accident will remain within the variations and limits of natural radioactivity and effects. (orig./HP) [de

  2. Análisis de series de tiempo para la predicción de los precios de la energía en la bolsa de Colombia

    Directory of Open Access Journals (Sweden)

    Cano Cano Jovan Alfonso

    2008-08-01

    Full Text Available Debido a la reestructuración del sector eléctrico colombiano,
    durante las dos últimas décadas, el comportamiento del precio de la
    energía eléctrica ha incrementado su volatilidad, reflejando el
    riesgo existente para los diferentes agentes que intervienen en el
    mercado. El objetivo de este artículo es presentar una metodología
    para la implementación de modelos de regresión, sobre la serie
    histórica de precios de bolsa de energía en Colombia. A medida que
    la cantidad de datos aumente, podrán desarrollarse modelos más
    amplios, que describan de forma adecuada comportamientos del
    mercado, que empleando las técnicas y la información disponible
    actualmente, no es posible identificar.

  3. Synthesis of the IRSN report related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor. Referral of the Permanent Group of Experts for nuclear reactors (GPR), examination of probabilistic level-2 safety studies (EPS 2) and severe accidents (AG) of the Flamanville reactor nr 3. Opinion related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor (FA3). Electronuclear reactors - EDF - Flamanville 3 EPR reactor. Severe accidents and probabilistic level 2 studies

    International Nuclear Information System (INIS)

    2015-01-01

    This document gathers several documents. The first one recalls the main arrangements implemented on the FA3 EPR reactor regarding accidents with core fusion, reports the analysis made by the IRSN about the sizing of these arrangements to reach a controlled status of the installation after a severe accident, regarding the probabilistic level-2 safety assessment, regarding the radiological impact of a severe accident on the population and on the environment, regarding those aimed at facing a total and long duration loss of electric power sources and cold sources, and about the situation of the reactor with respect to WENRA positions on severe accidents for new reactors. The second document is a letter sent by the ASN to the Permanent Group of Experts for nuclear reactors (GPR) to address probabilistic level-2 safety studies (EPS2) and severe accidents for the Flamanville 3 reactor. The third one reports the opinion of the GPR on these both issues and proposes a set of recommendations. The next document is a letter sent by the ASN to the Flamanville 3 project manager at EDF which recalls the related objectives, the ASN opinion on the implemented arrangements for severe accidents (de-pressurization of the primary circuit, management of hydrogen-related risks, corium recovery and cooling outside the vessel, limitation of vapour explosion risks outside the vessel, heat evacuation system, containment enclosure, management of the risk of a return to criticality), to face a total and long duration loss of electricity sources and cold sources, and other aspects addressed in the IRSN analysis. Requests and remarks formulated by the ASN are provided in an appendix to this last document

  4. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  5. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

    1987-01-01

    Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

  6. A Conceptual Study on a Supercritical CO_2-cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Yu, Hwanyeal; Hartanto, Donny; Kim, Yonghee

    2014-01-01

    A Micro Modular Reactor (MMR) using Supercritical-CO_2 (S-CO_2) as coolant has been investigated from the neutronics perspective. The MMR is designed to be transportable so it can reach the remote areas. The thermal power of the reactor is 36.2 M Wth. The size of the active core is limited to 1.2 m length and 93.16 cm width. The size of whole core is 2.8 m length and 166.9 cm width. The reactor lifetime design target is 20 years. To maximize the fuel volume fraction in the core, high density uranium nitride UN"1"5 was used. The PbO/MgO reflector was also utilized to improve the neutron economy. The S-CO_2 is chosen as the coolant because it offers a higher thermal efficiency. In this study, neutronics calculations and depletion using McCARD Monte Carlo code has been done to determine the lifetime and behavior of the core. Several important safety parameters such as Control Rod worth, Doppler reactivity coefficients and coolant void reactivity coefficient have also been analyzed. (author)

  7. Comparison of the N Reactor and Ignalina Unit No. 2 Level 1 Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Coles, G.A.; McKay, S.L.

    1995-06-01

    A multilateral team recently completed a full-scope Level 1 Probabilistic Safety Assessment (PSA) on the Ignalina Unit No. 2 reactor plant in Lithuania. This allows comparison of results to those of the PSA for the U.S. Department of Energy's (DOE) N Reactor. The N Reactor, although unique as a Western design, has similarities to Eastern European and Soviet graphite block reactors

  8. CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

    International Nuclear Information System (INIS)

    Kotas, J.F.; Stroh, K.R.

    1983-01-01

    The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power

  9. Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

    2010-01-01

    In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)

  10. Programa Bolsa Família: a interface entre a atuação profissional e o direito humano a alimentação adequada The "Bolsa Família" family grant scheme: the interface between professional practice and the human right to adequate food and nutrition

    Directory of Open Access Journals (Sweden)

    Camila Irigonhé Ramos

    2012-08-01

    Full Text Available O Direito Humano à Alimentação Adequada deve ser garantido através de políticas públicas de Segurança Alimentar e Nutricional (SAN. Nesse contexto está inserido o Programa Bolsa Família (PBF, que, além da transferência de renda, visa a garantia de acesso aos direitos sociais básicos. Este estudo objetiva analisar a operacionalização do PBF e, consequentemente, o entendimento dos profissionais de saúde a respeito do programa, enquanto eixo estruturante da política pública de SAN. Para isso, realizou-se entrevistas semiestruturadas com trabalhadores da atenção primária, envolvidos diretamente, tanto com o PBF, quanto com as famílias que recebem este beneficio. Ao final do estudo, foi possível evidenciar a importância da formação dos profissionais que atuam nessa área, pois, ao desconectar a realidade social em que os beneficiários estão inseridos, dos objetivos do programa, colabora-se para a simples mecanização dessas práticas. Nesse sentido, aponta-se que os profissionais de saúde precisam entender as proposições do programa como estratégias político-sociais, as quais, para além do alívio imediato, visam a superação dos problemas relacionados à pobreza e à fome.The Human Right to Adequate Nutrition must be ensured through the public policies included in SAN, namely the Food and Nutritional Security campaign. Besides the income transfer geared to ensuring access to basic social rights, the "Bolsa Família" Program (PBF is included in this context. This study seeks to analyze the operational aspects of the PBF and also ascertain whether or not the health professionals see the program as a core element of the SAN public policy. With this in mind, semi-structured interviews were conducted with primary healthcare workers involved directly both with the PBF and with the families who receive this benefit. By the end of the study, it was possible to perceive the importance of training health professionals who

  11. Preliminary Design of S-CO2 Brayton Cycle for KAIST Micro Modular Reactor

    International Nuclear Information System (INIS)

    Kim, Seong Gu; Kim, Min Gil; Bae, Seong Jun; Lee, Jeong Ik

    2013-01-01

    This paper suggests a complete modular reactor with an innovative concept of reactor cooling by using a supercritical carbon dioxide directly. Authors propose the supercritical CO 2 Brayton cycle (S-CO 2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the core and PCU in one vessel for the full modularization. This study suggests a conceptual design of small modular reactor including PCU which is named as KAIST Micro Modular Reactor (MMR). As a part of ongoing research of conceptual design of KAIST MMR, preliminary design of power generation cycle was performed in this study. Since the targets of MMR are full modularization of a reactor system with S-CO 2 coolant, authors selected a simple recuperated S-CO 2 Brayton cycle as a power conversion system for KAIST MMR. The size of components of the S-CO 2 cycle is much smaller than existing helium Brayton cycle and steam Rankine cycle, and whole power conversion system can be contained with core and safety system in one containment vessel. From the investigation of the power conversion cycle, recompressing recuperated cycle showed higher efficiency than the simple recuperated cycle. However the volume of heat exchanger for recompressing cycle is too large so more space will be occupied by heat exchanger in the recompressing cycle than the simple recuperated cycle. Thus, authors consider that the simple recuperated cycle is more suitable for MMR. More research for the KAIST MMR will be followed in the future and detailed information of reactor core and safety system will be developed down the road. More refined cycle layout and design of turbomachinery and heat exchanger will be performed in the future study

  12. Severe accident analysis for level 2 PSA of SMART reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jin Yong; Lee, Jeong Hun; Kim, Jong Uk; Yoo, Tae Geun; Chung, Soon Il; Kim, Min Gi [FNC Technology Co., Seoul (Korea, Republic of)

    2010-12-15

    The objectives of this study are to produce data for level 2 PSA and evaluation results of severe accident by analyzing severe accident sequence of transient events, producing fault tree of containment systems and evaluating direct containment heating of the SMART. In this project, severe accident analysis results were produced for general transient, loss of feedwater, station blackout, and steam line break events, and based on the results, design safety of SMART was verified. Also, direct containment heating phenomenon of the SMART was evaluated using TCE methodology. For level 2 PSA, fault tree of the containment isolation system, reactor cavity flooding system, plant chilled water system, and reactor containment building HVAC system was produced and analyzed

  13. Recent advances in the utilization and the irradiation technology of the refurbished BR2 reactor

    International Nuclear Information System (INIS)

    Dekeyser, J.; Benoit, P.; Decloedt, C.; Pouleur, Y.; Verwimp, A.; Weber, M.; Vankeerberghen, M.; Ponsard, B.

    1999-01-01

    Operation and utilization of the materials testing reactor BR2 at the Belgian Nuclear Research Centre (SCK·CEN) has since its start in 1963 always followed closely the needs and developments of nuclear technology. In particular, a multitude of irradiation experiments have been carried out for most types of nuclear power reactors, existing or under design. Since the early 1990s and increased focus was directed towards more specific irradiation testing needs for light water reactor fuels and materials, although other areas of utilization continued as well (e.g. fusion reactor materials, safety research, ...), including also the growing activities of radioisotope production and silicon doping. An important milestone was the decision in 1994 to implement a comprehensive refurbishment programme for the BR2 reactor and plant installations. The scope of this programme comprised very substantial studies and hardware interventions, which have been completed in early 1997 within planning and budget. Directly connected to this strategic decision for reactor refurbishment was the reinforcement of our efforts to requalify and upgrade the existing irradiation facilities and to develop advanced devices in BR2 to support emerging programs in the following fields: - LWR pressure vessel steel, - LWR irradiation assisted stress corrosion cracking (IASCC), - reliability and safety of high-burnup LWR fuel, - fusion reactor materials and blanket components, - fast neutron reactor fuels and actinide burning, - extension and diversification of radioisotope production. The paper highlights these advances in the areas of BR2 utilisation and the ongoing development activities for the required new generation of irradiations devices. (author)

  14. An improved thermal-hydraulic modeling of the Jules Horowitz Reactor using the CATHARE2 system code

    Energy Technology Data Exchange (ETDEWEB)

    Pegonen, R., E-mail: pegonen@kth.se [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden); Bourdon, S.; Gonnier, C. [CEA, DEN, DER, SRJH, CEA Cadarache, 13108 Saint-Paul-lez-Durance Cedex (France); Anglart, H. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden)

    2017-01-15

    Highlights: • An improved thermal-hydraulic modeling of the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during loss of flow accident. • The heat exchanger approach gives more realistic and less conservative results. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support current and future nuclear reactor designs. The reactor is under construction at the CEA Cadarache research center in southern France and is expected to achieve first criticality at the end of this decade. This paper presents an improved thermal-hydraulic modeling of the reactor using solely CATHARE2 system code. Up to now, the CATHARE2 code was simulating the full reactor with a simplified approach for the core and the boundary conditions were transferred into the three-dimensional FLICA4 core simulation. A new more realistic methodology is utilized to analyze the thermal-hydraulic simulation of the reactor during a loss of flow accident.

  15. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  16. Proceedings of 2. Yugoslav symposium on reactor physics, Part 2, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 2, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-07-01

    This Volume 2 of the Proceedings of 2. Yugoslav symposium on reactor physics includes eight papers dealing with the following topics: method for measuring high anti reactivities of a reactor system; integration method for thermal reaction rate calculation; Determination of initial core configuration for BHWR-200 MWe; safety shutdowns and failures of the RA reactor equipment; determining the reactivity of absorption rods; measurements of thermal and fast neutron fluxes at the TRIGA reactor and other measurements during operation of the TRIGA reactor; mathematical modelling of the reactor safety; review of problems and methods for radiation risk assessment in the environment of a nuclear power plant.

  17. Práticas discursivas e modos de subjetivação de mulheres beneficiárias do Programa Bolsa Família (PBF) em contextos rurais. O caso da Zona da Mata Pernambucana

    OpenAIRE

    MUNOZ, Claudio Baradit

    2016-01-01

    O presente estudo tem por objetivo analisar as práticas discursivas que constituem os modos de subjetivação de mulheres beneficiárias do Programa Bolsa Família (PBF) em contexto rural. Para isto será estudado o caso da Zona da Mata de Pernambuco. A metodologia qualitativa consiste na análise crítica do discurso. Os dados foram obtidos através de entrevistas semiestruturadas de seis mulheres. A fundamentação teórica é baseada no enfoque da governamentalidade, nas críticas feministas ao PBF e n...

  18. Las chicas en PISA y el marcado de casillas. Un examen de la perspectiva de los estudiantes sobre las pruebas PISA

    Directory of Open Access Journals (Sweden)

    Gerry Mac Ruairc

    2013-01-01

    Full Text Available El artículo se mueve desde el análisis a nivel macro a la perspectiva de los estudiantes en las pruebas PISA. Mientras que está empíricamente bien establecido el alto nivel de correlación entre el nivel educativo y el nivel socioeconómico de los estudiantes, en este estudio el autor pretende analizar cómo los estudiantes socio-económicamente desfavorecidos reaccionan a las pruebas y participan en el proceso. Para ello, es importante tener en cuenta los puntos de vista de los propios estudiantes. Al examinar los puntos de vista de los estudiantes en las pruebas de PISA en un estudio de caso, el autor ofrece una visión de cómo un grupo de chicas de clase trabajadora, procedente de una escuela de un área urbana desfavorecida, experimentó en la evaluación PISA (2009 en la República de Irlanda. El análisis temático de las entrevistas y las transcripciones a los grupos focales revelaron dos cuestiones: para la mayoría de los estudiantes, pero especialmente para aquellos con necesidades educativas especiales, se sintieron estresados por el contenido y la dificultad de los ítems; por su parte, los estudiantes se limitaban simplemente a los requisitos exigidos de completar la prueba en el tiempo establecido, con sus implicaciones para su validez. Se concluye con la necesidad de un enfoque más proactivo de apoyo a los estudiantes y un modelo más matizado de la evaluación en las futuras pruebas PISA para tener en cuenta las diferencias de clase social.

  19. Problems of nuclear reactor safety. Vol. 2

    International Nuclear Information System (INIS)

    Goncharov, L.A.

    1995-01-01

    Theses of proceedings of the 9 Topical Meeting on problems of nuclear power plant safety are presented. Reports include results of neutron-physical experiments carried out for reactor safety justification. Concepts of advanced reactors with improved safety are considered. Results of researches on fuel cycles are given too

  20. The Effect Of Beryllium Interaction With Fast Neutrons On the Reactivity Of ETRR-2 Research Reactor

    International Nuclear Information System (INIS)

    Aziz, M.; El Messiry, A.M.

    2000-01-01

    The effect of beryllium interactions with fast neutrons is studied for Etrr 2 research reactors. Isotope build up inside beryllium blocks is calculated under different irradiation times. a new model for the Etrr 2 research reactor is designed using MCNP code to calculate the reactivity and flux change of the reactor due to beryllium poison

  1. Application of the axial tomography computed for the detection of bags of dampness in dry wood of Gmelina arborea (Roxb.); Aplicacion de la tomografia axial computarizada para la deteccion de bolsas de humedad en madera seca de Gmelina arborea (Roxb.)

    Energy Technology Data Exchange (ETDEWEB)

    Moya R, Roger; Munoz A, Freddy [Inst. Tecnologico de Costa Rica, Escuela de Ingenieria Forestal, Apdo. 159-7050, Cartago 7050(Costa Rica); Escalante, Ivan [Clinica Santa Fe, San Jose (Costa Rica)

    2006-07-01

    Gmelina arborea (Roxb.) is widely used for commercial reforestation in Costa Rica due to its excellent growth rate and productivity. However, during the lumber drying process, the wooden boards show non-uniform values of final moisture content (MC). The low uniformity in final MC is caused by the presence of wet pockets, originated during the growing process of the tree. The regions with wet pockets present zones with a high MC, which are hard to detect with traditional methods for MC measurements during the wood drying process. It is possible to detect and to set the limits of the presence of wet wood in Gmelina arborea boards using scanning computed tomography (CT-scanning), a technique applied in medical diagnostic. A board with wet pockets is shown in the CT-scanning images in clear color and with low values of the Hounsfield Unit (HU) or CT number. When these values were transformed to wood density, it was determined that wet pockets were in a density of around 190 kg/m{sup 3}, a value higher than normal wood. Also, it was possible to observe growth tree rings in the CT-scanning images, an important feature for dendrochronological research. The obtained results allowed showing that it is possible to apply this technique in the process of lumber production, to detect the zones with high MC in kiln dried Gmelina arborea wood. (author) [Spanish] Gmelina arborea (Roxb.) es muy utilizado para la reforestacion comercial en Costa Rica debido a su excelente tasa de crecimiento y la productividad. Sin embargo, durante el proceso de secado, las tablas de madera muestran uniformidad en los valores finales de contenido de humedad (MC). La escasa uniformidad final de MC es causado por la presencia de bolsas humedas, se origino durante el proceso de crecimiento del arbol. Las regiones que presentan bolsas de humedad con un alto MC son dificiles de detectar con los metodos tradicionales de mediciones de MC durante el proceso de secado de madera. Se muestra que es posible

  2. Economics and utilization of thorium in nuclear reactors. Technical annexes 1 and 2

    International Nuclear Information System (INIS)

    1978-05-01

    An assessment of the impact of utilizing the 233 U/thorium fuel cycle in the U.S. nuclear economy is strongly dependent upon several decisions involving nuclear energy policy. These decisions include: (1) to recycle or not recycle fissile material; (2) if fissile material is recycled, to recycle plutonium, 233 U, or both; and (3) to deploy or not to deploy advanced reactor designs such as Fast Breeder Reactors (FBR's), High Temperature Gas Reactors (HTGR's), and Canadian Deuterium Uranium Reactors (CANDU's). This report examines the role of thorium in the context of the above policy decisions while focusing special attention on economics and resource utilization

  3. Variables antropométricas, hábitos y dietas alimentarias en adolescentes y jóvenes: diferencias en función del sexo

    Directory of Open Access Journals (Sweden)

    Carmen Maganto

    Full Text Available Resumen El estudio tuvo como objetivo analizar las diferencias entre sexos en variables antropométricas (reales, percibidas y deseadas, en hábitos alimentarios, y en el uso de dietas alimentarias. Los participantes fueron 1.075 adolescentes y jóvenes de 14 a 25 años (49.9 % varones, 50.1 % mujeres. Con un diseño descriptivo y comparativo, se administraron tres instrumentos de evaluación. Los resultados confirman muchas diferencias significativas entre sexos. En variables antropométricas las chicas se perciben más obesas de lo que están y desean estar más delgadas; los chicos se perciben igual o más delgados de lo que están y desean tener un volumen corporal superior. Los chicos desean tener un Índice de Masa Corporal (IMC superior y las chicas inferior. Las chicas obtienen puntuaciones significativamente superiores en hábitos alimentarios, aunque los chicos perciben que tienen una alimentación más equilibrada. Las chicas han realizado más dietas y creen necesitarlas más. Las razones para engordar en los chicos son biológicas y en las chicas hábitos alimentarios inadecuados. Las chicas realizan más dietas tanto saludables como no recomendables. Las razones para comenzar una dieta son en las chicas la imagen corporal y en los chicos ser aceptado por los iguales. El abandono de las dietas los chicos lo atribuyen a la dieta y las chicas a sí mismas. El estudio aporta datos relevantes para el diseño de programas preventivos y/o de tratamiento con adolescentes/jóvenes con problemas alimentarios, bien por alteraciones de la imagen corporal, hábitos alimentarios inadecuados y/o por el uso indebido de dietas alimentarias.

  4. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    International Nuclear Information System (INIS)

    Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Pal, S.; Mishra, G.P.

    2012-01-01

    CERMET fuel with either PuO 2 or enriched UO 2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO 2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO 2 , U–25 wt%UO 2 and U–30 wt%UO 2 . It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO 2 compositions.

  5. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    Science.gov (United States)

    Sinha, V. P.; Hegde, P. V.; Prasad, G. J.; Pal, S.; Mishra, G. P.

    2012-08-01

    CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

  6. Utilization of the SLOWPOKE-2 research reactor

    International Nuclear Information System (INIS)

    Lalor, G.C.

    2001-01-01

    SLOWPOKEs are typically low power research reactors that have a limited number of applications. However, a significant range of NAA can be performed with such reactors. This paper describes a SLOWPOKE-based NAA program that is performing a valuable series of studies in Jamaica, including geological mapping and pollution assessment. (author)

  7. Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

  8. Measurements at the RA Reactor related to the VISA-2 project - Part 1, Start-up of the RA reactor and measurement of new RA reactor core parameters

    International Nuclear Information System (INIS)

    Markovic, H.

    1962-07-01

    The objective of the measurements was determining the neutron flux in the RA reactor core. Since the number of fuel channels is increased from 56 to 68 within the VISA-2 project, it was necessary to attain criticality of the RA reactor and measure the neutron flux properties. The 'program of RA reactor start-up' has been prepared separately and it is included in this report. Measurements were divided in two phases. First phase was measuring of the neutron flux after the criticality was achieved but at zero power. During phase two measurements were repeated at several power levels, at equilibrium xenon poisoning. This report includes experimental data of flux distributions and absolute values of the thermal and fast neutron flux in the RA reactor experimental channels and values of cadmium ratio for determining the neutron epithermal flux. Data related to calibration of regulatory rods for cold un poisoned core are included [sr

  9. Reactor limitation system improves the safety and availability of the Angra 2 nuclear power plant

    International Nuclear Information System (INIS)

    Souza Mendes, J.E. de

    1987-01-01

    Beyond the classic Reactor Protection System and Reactor Control System, nuclear plant Angra 2 has a third system called Reactor Limitation System which combines the intelligence features of the control systems with the high reliability of the protection systems. In determined events, which are not controlled by the control system (e.g.: load rejection, failure of one main reactor coolant pump), the Reactor Limitation System actuates automatically in order to lead the plant to a safe operating condition and so it avoids the actuation of the Reactor Protection System and consequently the reactor trip. This increases safety and availability of the plant and reduces component stresses. After the safe operating condition is reached, the process guidance automatically returns to the control systems. (Author) [pt

  10. CO_2 capture with solid sorbent: CFD model of an innovative reactor concept

    International Nuclear Information System (INIS)

    Barelli, L.; Bidini, G.; Gallorini, F.

    2016-01-01

    Highlights: • A new reactor solution based on rotating fixed beds was presented. • The preliminary design of the reactor was approached. • A CFD model of the reactor, including CO_2 capture kinetic, was developed. • The CFD model is validated with experimental results. • Sorbent exploitation increasing is possible thanks to the new reactor. - Abstract: In future decarbonization scenarios, CCS with particular reference to post-combustion technologies will be an important option also for energy intensive industries. Nevertheless, today CCS systems are rarely installed due to high energy and cost penalties of current technology based on chemical scrubbing with amine solvent. Therefore, innovative solutions based on new/optimized solvents, sorbents, membranes and new process designs, are R&D priorities. Regarding the CO_2 capture through solid sorbents, a new reactor solution based on rotating fixed beds is presented in this paper. In order to design the innovative system, a suitable CFD model was developed considering also the kinetic capture process. The model was validated with experimental results obtained by the authors in previous research activities, showing a potential reduction of energy penalties respect to current technologies. In the future, the model will be used to identify the control logic of the innovative reactor in order to verify improvements in terms of sorbent exploitation and reduction of system energy consumption.

  11. MULTI-LOOP CONTROL DESIGN IN MULTIVARIABLE (2X2 CONTINUOUS STIRRED TANK REACTOR

    Directory of Open Access Journals (Sweden)

    Abdul Wahid

    2015-06-01

    Full Text Available With this study, the design and tuning of multi-loop for multivariable (2x2 CSTR will be made in order to achieve optimum CSTR control performance. This study used Bequette model reactor and MATLAB software and is expected to be able to cope with disturbances in the reactor so that the reactor system is able to stabilize quickly despite the distractions. In this study, the design will be made using multi-loop approach, along with PI controller as the next step. Then, BLT and auto-tune tuning method will be used in PI controller and given disturbances to both of tuning method. The controller performances are then compared. Results of the study are then analyzed for discussions and conclusions. Results from this study have shown that in terms of disturbance rejection, BLT is better than auto-tune based on comparison between both of controller performances. For IAE for the case of temperature, BLT is 30% better than auto-tune, but it is almost the same for the case of concentration. For settling time for the case of concentration, BLT is 30% better than auto-tune, and for the case of temperature, BLT is 18% better than auto-tune. For rise time for the case of concentration and temperature, BLT is 30% better than auto-tune.

  12. Operating reactors licensing actions summary. Volume 5, No. 2

    International Nuclear Information System (INIS)

    1985-04-01

    The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published primarily for internal NRC use in managing the Operating Reactors Licensing Actions Program

  13. Tests of Neutron Spectrum Calculations with the Help of Foil Measurements in a D{sub 2}O and in an H{sub 2}O-Moderated Reactor and in Reactor Shields of Concrete an Iron

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, R; Aalto, E

    1964-09-15

    Foil measurements covering the fast, epithermal and thermal neutron energy regions have been made in the centre of the Swedish D{sub 2}O-moderated reactor R1, in the pool reactor R2-0, and in different positions in reactor shields of iron, magnetite concrete and ordinary concrete. Neutron spectra have also been calculated for most of these positions, often with the help of a numerical integration of the Boltzmann equation. The measurements and the calculated spectra are presented.

  14. Microflow photochemistry: UVC-induced [2 + 2]-photoadditions to furanone in a microcapillary reactor

    Directory of Open Access Journals (Sweden)

    Sylvestre Bachollet

    2013-10-01

    Full Text Available [2 + 2]-Cycloadditions of cyclopentene and 2,3-dimethylbut-2-ene to furanone were investigated under continuous-flow conditions. Irradiations were conducted in a FEP-microcapillary module which was placed in a Rayonet chamber photoreactor equipped with low wattage UVC-lamps. Conversion rates and isolated yields were compared to analogue batch reactions in a quartz test tube. In all cases examined, the microcapillary reactor furnished faster conversions and improved product qualities.

  15. A simulation Model of the Reactor Hall Ventilation and air Conditioning Systems of ETRR-2

    International Nuclear Information System (INIS)

    Abd El-Rahman, M.F.

    2004-01-01

    Although the conceptual design for any system differs from one designer to another. each of them aims to achieve the function of the system required. the ventilation and air conditioning system of reactors hall is one of those systems that really differs but always dose its function for which it is designed. thus, ventilation and air conditioning in some reactor hall constitute only one system whereas in some other ones, they are separate systems. the Egypt Research Reactor-2 (ETRR-2)represents the second type. most studies conducted on ventilation and air conditioning simulation models either in traditional building or for research rectors show that those models were not designed similarly to the model of the hall of ETRR-2 in which ventilation and air conditioning constitute two separate systems.besides, those studies experimented on ventilation and air conditioning simulation models of reactor building predict the temperature and humidity inside these buildings at certain outside condition and it is difficult to predict when the outside conditions are changed . also those studies do not discuss the influences of reactor power changes. therefore, the present work deals with a computational study backed by infield experimental measurements of the performance of the ventilation and air conditioning systems of reactor hall during normal operation at different outside conditions as well as at different levels of reactor power

  16. Thermal design of heat-exchangeable reactors using a dry-sorbent CO2 capture multi-step process

    International Nuclear Information System (INIS)

    Moon, Hokyu; Yoo, Hoanju; Seo, Hwimin; Park, Yong-Ki; Cho, Hyung Hee

    2015-01-01

    The present study proposes a multi-stage CO 2 capture process that incorporates heat-exchangeable fluidized-bed reactors. For continuous multi-stage heat exchange, three dry regenerable sorbents: K 2 CO 3 , MgO, and CaO, were used to create a three-stage temperature-dependent reaction chain for CO 2 capture, corresponding to low (50–150 °C), middle (350–650 °C), and high (750–900 °C) temperature stages, respectively. Heat from carbonation in the high and middle temperature stages was used for regeneration for the middle and low temperature stages. The feasibility of this process is depending on the heat-transfer performance of the heat-exchangeable fluidized bed reactors as the focus of this study. The three-stage CO 2 capture process for a 60 Nm 3 /h CO 2 flow rate required a reactor area of 0.129 and 0.130 m 2 for heat exchange between the mid-temperature carbonation and low-temperature regeneration stages and between the high-temperature carbonation and mid-temperature regeneration stages, respectively. The reactor diameter was selected to provide dense fluidization conditions for each bed with respect to the desired flow rate. The flow characteristics and energy balance of the reactors were confirmed using computational fluid dynamics and thermodynamic analysis, respectively. - Highlights: • CO 2 capture process is proposed using a multi-stage process. • Reactor design is conducted considering heat exchangeable scheme. • Reactor surface is designed by heat transfer characteristics of fluidized bed

  17. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

  18. Core design calculations of IRIS reactor using modified CORD-2 code package

    International Nuclear Information System (INIS)

    Pevec, D.; Grgic, D.; Jecmenica, R.; Petrovic, B.

    2002-01-01

    Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)

  19. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  20. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  1. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  2. BR2 reactor: medical and industrial applications

    International Nuclear Information System (INIS)

    Ponsard, B.

    2005-01-01

    The radioisotopes are produced for various applications in the nuclear medicine (diagnostic, therapy, palliation of metastatic bone pain), industry (radiography of welds, ...), agriculture (radiotracers, ...) and basic research. Due to the availability of high neutron fluxes (thermal neutron flux up to 10 15 n/cm 2 .s), the BR2 reactor is considered as a major facility through its contribution for a continuous supply of products such 99 Mo ( 99 mTc), 131 I, 133 Xe, 192 Ir, 186 Re, 153 Sm, 90 Y, 32 P, 188 W ( 188 Re), 203 Hg, 82 Br, 41 Ar, 125 I, 177 Lu, 89 Sr, 60 Co, 169 Yb, 147 Nd, and others. Neutron Transmutation Doped (NTD) silicon is produced for the semiconductor industry in the SIDONIE (Silicon Doping by Neutron Irradiation Experiment) facility, which is designed to continuously rotate and traverse the silicon through the neutron flux. These combined movements produce exceptional dopant homogeneity in batches of silicon measuring 4 and 5-inches in diameter by up to 750 mm in length. The main objectives of work performed were to provide a reliable and qualitative supply of radioisotopes and NTD-silicon to the customers in accordance with a quality system that has been certified to the requirements of the EN ISO 9001: 2000. This new Quality System Certificate has been obtained in November 2003 for the Production of radioisotopes for medical and industrial applications and the Production of Neutron Transmutation Doped (NTD) Silicon in the BR2 reactor

  3. Possible future roles for fast breeder reactors Part 1 and 2

    International Nuclear Information System (INIS)

    1978-06-01

    Part 1. The Fast Breeder Reactor (in particular in its sodium cooled version) has been steadily developed in the Community. This report attempts to quantify the advantages of this system in terms of fossil energy and uranium savings in the medium/long term as well as to examine some long term economic implications. The methodology of comparing scenarios, not individual reactor systems is followed. These scenarios have been chosen taking into account a range of assumptions concerning Community energy demand growth, fossil energy and uranium availability and technological capabilities. Part 2. The fast breeder reactor (FBR), particularly its sodium-cooled form (LMFBR) has been under development in the Community for many years. Industrial enterprises dedicated to its commercialisation have been formed and long range plans for its industrial utilisation are being formulated. The value of breeder reactors from the point of view of minimising Community fuel requirements has been discussed in Part I of this report (1). In Part II the consequences of delaying their introduction, and the demands placed upon the recycle industry by the introduction of fast reactors of different characteristics, using the Community electricity demand scenarios developed for Part I, are discussed. In addition comments are provided upon the effect of FBR introduction on the size of plutonium stocks

  4. As contribuições do Programa Institucional de Bolsas de Iniciação à Docência para a formação docente

    Directory of Open Access Journals (Sweden)

    Gleicy Calhau Gomes

    2012-12-01

    Full Text Available Este artigo trata da formação docente para o ensino fundamental e tem por objetivo investigar as contribuições e desafios do Programa Institucional de Bolsas de Iniciação à Docência (PIBID. Realizou-se a pesquisa à partir da metodologia de pesquisa qualitativa segundo a perspectiva fenomenológica, com enfoque na pesquisa-ação. Dentre os sujeitos pesquisados, os principais foram os bolsistas e egressos do PIBID de Pedagogia da UNEMAT- campus Universitário de Sinop-MT. Conclui-se que a formação dos acadêmicos/bolsistas, e a parceria entre universidade/escola, contribui com as práticas educativas idealizadas na escola pública, e com uma formação de mais qualidade para estes bolsistas. Palavras-chave: educação; formação docente; PIBID; práticas educativas; pesquisa-ação. 

  5. Programa Bolsa Família e segurança alimentar e nutricional no Brasil: revisão crítica da literatura The Bolsa Família cash transfer program and food and nutrition security in Brazil: a critical review of the literature

    Directory of Open Access Journals (Sweden)

    Rosângela Minardi Mitre Cotta

    2013-01-01

    Full Text Available OBJETIVO: Revisar criticamente os estudos que avaliaram os impactos do Programa Bolsa Família (PBF na promoção da segurança alimentar e nutricional no Brasil. MÉTODOS: Foram consultadas as bases de dados Biblioteca Cochrane, LILACS, Medline e SciELO, bem como os portais de organizações públicas. Foram selecionados os estudos que utilizaram dados primários e excluídos estudos baseados em dados secundários, artigos de revisão, estudos que não permitiram estabelecer uma associação entre PBF e segurança alimentar e nutricional, bem como os estudos que avaliaram a segurança do alimento no que se refere apenas à qualidade sanitária. RESULTADOS Foram selecionados 10 estudos, dos quais cinco concluíram que o PBF teve um impacto positivo na segurança alimentar e nutricional das famílias beneficiárias. Entretanto, três estudos constataram um aumento do consumo de alimentos de maior densidade calórica e baixo valor nutritivo. Essa mudança no hábito alimentar é um fator de risco para o desenvolvimento do sobrepeso, obesidade e das doenças crônicas não transmissíveis. CONCLUSÕES: A garantia de segurança alimentar e nutricional exige programas que contemplem tanto o combate à desnutrição quanto ao sobrepeso e à obesidade. Programas de distribuição de renda, como o PBF, podem contribuir mais efetivamente para o bem-estar nutricional dos beneficiários quando combinados com outros tipos de intervenções, como ações de promoção de alimentação saudável.OBJECTIVE: To critically review studies evaluating the impact of Bolsa Família (PBF, a federal cash transfer program, for food and nutrition security in Brazil. METHODS: The Cochrane Library, LILACS, Medline and SciELO databases were searched, as well as public organization websites. All studies based on primary data were selected. The following were excluded: studies using secondary data, review articles, studies that did now allow the establishment of associations

  6. A regression approach for Zircaloy-2 in-reactor creep constitutive equations

    International Nuclear Information System (INIS)

    Yung Liu, Y.; Bement, A.L.

    1977-01-01

    In this paper the methodology of multiple regressions as applied to Zircaloy-2 in-reactor creep data analysis and construction of constitutive equation are illustrated. While the resulting constitutive equation can be used in creep analysis of in-reactor Zircaloy structural components, the methodology itself is entirely general and can be applied to any creep data analysis. The promising aspects of multiple regression creep data analysis are briefly outlined as follows: (1) When there are more than one variable involved, there is no need to make the assumption that each variable affects the response independently. No separate normalizations are required either and the estimation of parameters is obtained by solving many simultaneous equations. The number of simultaneous equations is equal to the number of data sets. (2) Regression statistics such as R 2 - and F-statistics provide measures of the significance of regression creep equation in correlating the overall data. The relative weights of each variable on the response can also be obtained. (3) Special regression techniques such as step-wise, ridge, and robust regressions and residual plots, etc., provide diagnostic tools for model selections. Multiple regression analysis performed on a set of carefully selected Zircaloy-2 in-reactor creep data leads to a model which provides excellent correlations for the data. (Auth.)

  7. Coolant radiolysis studies in the high temperature, fuelled U-2 loop in the NRU reactor

    International Nuclear Information System (INIS)

    Elliot, A.J.; Stuart, C.R.

    2008-06-01

    An understanding of the radiolysis-induced chemistry in the coolant water of nuclear reactors is an important key to the understanding of materials integrity issues in reactor coolant systems. Significant materials and chemistry issues have emerged in Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR) and CANDU reactors that have required a detailed understanding of the radiation chemistry of the coolant. For each reactor type, specific computer radiolysis models have been developed to gain insight into radiolysis processes and to make chemistry control adjustments to address the particular issue. In this respect, modelling the radiolysis chemistry has been successful enough to allow progress to be made. This report contains a description of the water radiolysis tests performed in the U-2 loop, NRU reactor in 1995, which measured the CHC under different physical conditions of the loop such as temperature, reactor power and steam quality. (author)

  8. Design and computational analysis of passive siphon breaker for 49-2 swimming pool reactor

    International Nuclear Information System (INIS)

    Yue Zhiting; Song Yunpeng; Liu Xingmin; Zou Yao; Wu Yuanyuan

    2014-01-01

    Based on safety considerations, a passive siphon breaker will be added to the primary cooling system of 49-2 Swimming Pool Reactor (SPR). With the breaker location determined, the capability of siphon breakers with diameters of 1.5 cm and 2.0 cm was calculated and analyzed respectively by RELAP5/MOD3.3 code. The results show that in the condition of large break loss of coolant accident these two sizes of siphon breakers are able to break the siphon phenomena, and maintain the pool water level above the reactor core when the reactor and the pump are shutdown. In the end, to be conservative, the siphon breaker with diameter of 2.0 cm is adopted. (authors)

  9. Efficient H2O2/CH3COOH oxidative desulfurization/denitrification of liquid fuels in sonochemical flow-reactors.

    Science.gov (United States)

    Calcio Gaudino, Emanuela; Carnaroglio, Diego; Boffa, Luisa; Cravotto, Giancarlo; Moreira, Elizabeth M; Nunes, Matheus A G; Dressler, Valderi L; Flores, Erico M M

    2014-01-01

    The oxidative desulfurization/denitrification of liquid fuels has been widely investigated as an alternative or complement to common catalytic hydrorefining. In this process, all oxidation reactions occur in the heterogeneous phase (the oil and the polar phase containing the oxidant) and therefore the optimization of mass and heat transfer is of crucial importance to enhancing the oxidation rate. This goal can be achieved by performing the reaction in suitable ultrasound (US) reactors. In fact, flow and loop US reactors stand out above classic batch US reactors thanks to their greater efficiency and flexibility as well as lower energy consumption. This paper describes an efficient sonochemical oxidation with H2O2/CH3COOH at flow rates ranging from 60 to 800 ml/min of both a model compound, dibenzotiophene (DBT), and of a mild hydro-treated diesel feedstock. Four different commercially available US loop reactors (single and multi-probe) were tested, two of which were developed in the authors' laboratory. Full DBT oxidation and efficient diesel feedstock desulfurization/denitrification were observed after the separation of the polar oxidized S/N-containing compounds (S≤5 ppmw, N≤1 ppmw). Our studies confirm that high-throughput US applications benefit greatly from flow-reactors. Copyright © 2013 Elsevier B.V. All rights reserved.

  10. Direct In Situ Quantification of HO2 from a Flow Reactor.

    Science.gov (United States)

    Brumfield, Brian; Sun, Wenting; Ju, Yiguang; Wysocki, Gerard

    2013-03-21

    The first direct in situ measurements of hydroperoxyl radical (HO2) at atmospheric pressure from the exit of a laminar flow reactor have been carried out using mid-infrared Faraday rotation spectroscopy. HO2 was generated by oxidation of dimethyl ether, a potential renewable biofuel with a simple molecular structure but rich low-temperature oxidation chemistry. On the basis of the results of nonlinear fitting of the experimental data to a theoretical spectroscopic model, the technique offers an estimated sensitivity of reactor exit temperature range of 398-673 K. Accurate in situ measurement of this species will aid in quantitative modeling of low-temperature and high-pressure combustion kinetics.

  11. Reactor science and technology: operation and control of reactors

    International Nuclear Information System (INIS)

    Qiu Junlong

    1994-01-01

    This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

  12. Status and perspective of development of cold moderators at the IBR-2 reactor

    International Nuclear Information System (INIS)

    Kulikov, S; Shabalin, E

    2012-01-01

    The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams nos. 7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams nos. 2-3 and for beams nos. 1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (∼3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

  13. Status and perspective of development of cold moderators at the IBR-2 reactor

    Science.gov (United States)

    Kulikov, S.; Shabalin, E.

    2012-03-01

    The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams #7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams #2-3 and for beams #1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (~3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

  14. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  15. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  16. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

  17. Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

    International Nuclear Information System (INIS)

    Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

    2003-01-01

    The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

  18. Further study on parameterization of reactor NAA: Pt. 2

    International Nuclear Information System (INIS)

    Tian Weizhi; Zhang Shuxin

    1989-01-01

    In the last paper, Ik 0 method was proposed for fission interference corrections. Another important kind of interferences in reator NAA is due to threshold reaction induced by reactor fast neutrons. In view of the increasing importance of this kind of interferences, and difficulties encountered in using the relative comparison method, a parameterized method has been introduced. Typical channels in heavy water reflector and No.2 horizontal channel of Heavy Water Research Reactor in the Insitute of Atomic Energy have been shown to have fast neutron energy distributions (E>4 MeV) close to primary fission neutron spectrum, by using multi-threshold detectors. On this basis, Ti foil is used as an 'instant fast neutron flux monitor' in parameterized corrections for threshold reaction interferences in the long irradiations. Constant values of φ f /φ s = 0.70 ± 0.02% have been obtained for No.2 rabbit channel. This value can be directly used for threshold reaction inference correction in the short irradiations

  19. EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

    Energy Technology Data Exchange (ETDEWEB)

    Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  20. The analysis for inventory of experimental reactor high temperature gas reactor type

    International Nuclear Information System (INIS)

    Sri Kuntjoro; Pande Made Udiyani

    2016-01-01

    Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

  1. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Memmott, Matthew; Boy, Guy; Charit, Indrajit; Manera, Annalisa; Downar, Thomas; Lee, John; Muldrow, Lycurgus; Upadhyaya, Belle; Hines, Wesley; Haghighat, Alierza

    2017-01-01

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project ''Integral Inherently Safe Light Water Reactors (I 2 S-LWR)''. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

  2. Continuous backfitting measures for the FRG-1 and FRG-2 research reactors

    International Nuclear Information System (INIS)

    Blom, K.H.; Falck, K.; Krull, W.

    1990-01-01

    The GKSS-Research Centre Geesthacht GmbH has been operating the research reactors FRG-1 and FRG-2 with power levels of 5 MW and 15 MW for 31 and 26 years respectively. Safe operation at full power levels over so many years with an average utilization between 180 d to 250 d per year is possible only with great efforts in modernization and upgrading of the research reactors. Emphasis has been placed on backfitting since around 1975. At that time within the Federal Republic of Germany many new guidelines, rules, ordinances, and standards in the field of (power) reactor safety were published. Much work has been done on the modernization of the FRG-1 and FRG-2 research reactors therefore within the last ten years. Work done within the last two years and presently underway includes: measures against water leakage through the concrete and along the beam tubes; repair of both cooling towers; modernization of the ventilation system; measures for fire protection; activities in water chemistry and water quality; installation of a double tubing for parts of the primary piping of the FRG-1; replacement of instrumentation, process control systems (operation and monitoring system) and alarm system; renewal of the emergency power supply; installation of internal lightning protection; installation of a cold neutron source; enrichment reduction for FRG-1. These efforts will continue to allow safe operation of our research reactors over their whole operational life

  3. TiO2-photocatalyzed As(III) oxidation in a fixed-bed, flow-through reactor.

    Science.gov (United States)

    Ferguson, Megan A; Hering, Janet G

    2006-07-01

    Compliance with the U.S. drinking water standard for arsenic (As) of 10 microg L(-1) is required in January 2006. This will necessitate implementation of treatment technologies for As removal by thousands of water suppliers. Although a variety of such technologies is available, most require preoxidation of As(III) to As(V) for efficient performance. Previous batch studies with illuminated TiO2 slurries have demonstrated that TiO2-photocatalyzed AS(III) oxidation occurs rapidly. This study examined reaction efficiency in a flow-through, fixed-bed reactor that provides a better model for treatment in practice. Glass beads were coated with mixed P25/sol gel TiO2 and employed in an upflow reactor irradiated from above. The reactor residence time, influent As(III) concentration, number of TiO2 coatings on the beads, solution matrix, and light source were varied to characterize this reaction and determine its feasibility for water treatment. Repeated usage of the same beads in multiple experiments or extended use was found to affect effluent As(V) concentrations but not the steady-state effluent As(III) concentration, which suggests that As(III) oxidation at the TiO2 surface undergoes dynamic sorption equilibration. Catalyst poisoning was not observed either from As(V) or from competitively adsorbing anions, although the higher steady-state effluent As(III) concentrations in synthetic groundwater compared to 5 mM NaNO3 indicated that competitive sorbates in the matrix partially hinder the reaction. A reactive transport model with rate constants proportional to incident light at each bead layer fit the experimental data well despite simplifying assumptions. TiO2-photocatalyzed oxidation of As(III) was also effective under natural sunlight. Limitations to the efficiency of As(III) oxidation in the fixed-bed reactor were attributable to constraints of the reactor geometry, which could be overcome by improved design. The fixed-bed TiO2 reactor offers an environmentally

  4. Exxon nuclear neutronics design methods for pressurized water reactors. Supplement 2

    International Nuclear Information System (INIS)

    Skogen, F.B.; Stout, R.B.

    1977-01-01

    Modifications to the Exxon Nuclear PWR neutronic design calculational methods are presented as well as the results obtained when these improved methods are compared to reactor measurements. The basic PWR design tools remain unchanged; i.e., the XPOSE code is used for generating the basic nuclear parameters, the PDQ-7 code is used for calculating reactivity and x-y power distributions, and the XTG code is used for three-dimensional analysis. The recent start-up experiences at D. C. Cook Unit 1 and H. B. Robinson Unit 2 have provided a significant increase in the data base supporting the current ENC PWR neutronic methods. The verification comparisons contained in the supplement include reactor measurements from D. C. Cook Unit 1, Cycle 2; H. B. Robinson Unit 2, Cycles 4 and 5; Palisades Cycle 2, and R. E. Ginna, Cycle 7

  5. Characterization of fuel distribution in the Three Mile Island Unit 2 (TMI-2) reactor system by neutron and gamma-ray dosimetry

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McNeece, J.P.; Kaiser, B.J.; McElroy, W.N.

    1984-01-01

    Neutron and gamma-ray dosimetry are being used for nondestructive assessment of the fuel distribution throughout the Three Mile Island Unit 2 (TMI-2) reactor core region and primary cooling system. The fuel content of TMI-2 makeup and purification Demineralizer A has been quantified with Si(Li) continuous gamma-ray spectrometry and solid-state track recorder (SSTR) neutron dosimetry. For fuel distribution characterization in the core region, results from SSTR neutron dosimetry exposures in the TMI-2 reactor cavity are presented. These SSTR results are consistent with the presence of a significant amount of fuel debris, equivalent to several fuel assemblies or more, lying at the bottom of the reactor vessel. (Auth.)

  6. Estimation of power feedback parameters of the IBR-2M reactor by square wave reactivity

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Popov, A.K.; Sumkhuu, D.

    2016-01-01

    Parameters of the IBR-2M reactor power feedback (PFB) are estimated based on the analysis of power transients caused by deliberate square wave reactivity when the pulsed reactor operates in the self-regulation mode. The PFB of the IBR-2M is described by three linear first-order differential equations. Two components of the PFB are responsible for the negative feedback and one, for the positive. The overall feedback is negative, i.e., it has a stabilizing effect for the operation of the reactor. The slowest negative component of the PFB is probably caused by heating of the fuel. Periodically repeated in the process of exploitation, estimation of the PFB parameters is one of the methods to ensure safety operation of the reactor. [ru

  7. Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 3 of the Proceedings of 2. Yugoslav symposium on reactor physics includes three papers describing the following: model for spatial synthesis of automated control system of the GCR type reactor; model for analysis of hydrodynamic processes at the BHWR type reactors; mathematical model for safety analysis of heavy water power reactor

  8. TOKMINA, Toroidal Magnetic Field Minimization for Tokamak Fusion Reactor. TOKMINA-2, Total Power for Tokamak Fusion Reactor

    International Nuclear Information System (INIS)

    Hatch, A.J.

    1975-01-01

    1 - Description of problem or function: TOKMINA finds the minimum magnetic field, Bm, required at the toroidal coil of a Tokamak type fusion reactor when the input is beta(ratio of plasma pressure to magnetic pressure), q(Kruskal-Shafranov plasma stability factor), and y(ratio of plasma radius to vacuum wall radius: rp/rw) and arrays of PT (total thermal power from both d-t and tritium breeding reactions), Pw (wall loading or power flux) and TB (thickness of blanket), following the method of Golovin, et al. TOKMINA2 finds the total power, PT, of such a fusion reactor, given a specified magnetic field, Bm, at the toroidal coil. 2 - Method of solution: TOKMINA: the aspect ratio(a) is minimized, giving a minimum value for Bm. TOKMINA2: a search is made for PT; the value of PT which minimizes Bm to the required value within 50 Gauss is chosen. 3 - Restrictions on the complexity of the problem: Input arrays presently are dimensioned at 20. This restriction can be overcome by changing a dimension card

  9. System Definition Document: Reactor Data Necessary for Modeling Plutonium Disposition in Catawba Nuclear Station Units 1 and 2

    International Nuclear Information System (INIS)

    Ellis, R.J.

    2000-01-01

    The US Department of Energy (USDOE) has contracted with Duke Engineering and Services, Cogema, Inc., and Stone and Webster (DCS) to provide mixed-oxide (MOX) fuel fabrication and reactor irradiation services in support of USDOE's mission to dispose of surplus weapons-grade plutonium. The nuclear station units currently identified as mission reactors for this project are Catawba Units 1 and 2 and McGuire Units 1 and 2. This report is specific to Catawba Nuclear Station Units 1 and 2, but the details and materials for the McGuire reactors are very similar. The purpose of this document is to present a complete set of data about the reactor materials and components to be used in modeling the Catawba reactors to predict reactor physics parameters for the Catawba site. Except where noted, Duke Power Company or DCS documents are the sources of these data. These data are being used with the ORNL computer code models of the DCS Catawba (and McGuire) pressurized-water reactors

  10. Report on the operation in 1973 of the FR 2 research reactor

    International Nuclear Information System (INIS)

    Moeller, I.; Steiger, W.

    1975-04-01

    Also in 1973, the heavy-water moderated research and testing reactor FR 2 was operated to schedule at 44 MW nominal power. Again, the availability of the plant was slightly improved. Experimental utilization through instrumented irradiation capsules strongly increased as compared to the previous year. Up to 16 capsule test rigs at a time were inserted in the reactor. As to the beam tube experiments, up to 13 experiments covering a total of 18 test rigs were conducted simultaneously at the 12 reasonably usable beam holes. At the beginning of the year all of the positions available were occupied by 5 loop experiments. Isotope production reached its highest value with a total of 2,372 irradiated capsules (1.3% more than the year before). Some remarkable figures characterized the year 1973: On August 16, 1973 ten years of full power operation at a nominal power of 12 and 44 MW, respectively, had been reached. On July 24, 1973 the 50,000th isotope irradiation was performed in the reactor and on December 26, 1973 a total energy release of 100,000 MWd was recorded. Moreover, the 125,000th visitor of the reactor was welcomed on December 6, 1973. (orig./UA) [de

  11. Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor building decontamination. Summary status report. Volume 2

    International Nuclear Information System (INIS)

    Doerge, D.H.; Miller, R.L.; Scotti, K.S.

    1986-05-01

    This document summarizes information relating to decontamination of the Three Mile Island Unit 2 (TMI-2) reactor building. The report covers activities for the period of June 1, 1979 through March 29, 1985. The data collected from activity reports, reactor containment entry records, and other sources were entered into a computerized data system which permits extraction/manipulation of specific information which can be used in planning for recovery from an accident similar to that experienced at TMI-2 on March 28, 1979. This report contains summaries of man-hours, manpower, and radiation exposures incurred during decontamination of the reactor building. Support activities conducted outside of radiation areas are excluded from the scope of this report. Computerized reports included in this document are: a chronological summary listing work performed relating to reactor building decontamination for the period specified; and summary reports for each major task during the period. Each task summary is listed in chronological order for zone entry and subtotaled for the number of personnel entries, exposures, and man-hours. Manually-assembled table summaries are included for: labor and exposures by department and labor and exposures by major activity

  12. Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

    Energy Technology Data Exchange (ETDEWEB)

    Dougherty, D.; Adams, J. W.

    1983-08-01

    Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation.

  13. Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

    International Nuclear Information System (INIS)

    Dougherty, D.; Adams, J.W.

    1983-08-01

    Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation

  14. MANEJO DE VÍA AÉREA NO DIFÍCIL. DESDE LA VENTILACIÓN CON BOLSA HASTA INTUBACIÓN ORO TRAQUEAL

    Directory of Open Access Journals (Sweden)

    Ramón Coloma, Dr.

    2017-09-01

    Full Text Available RESUMEN: El manejo de la vía aérea no difícil es una de las habilidades que todo médico debiera dominar. Para ello se requiere el conocimiento de conceptos básicos tanto anatómicos como fisiológicos, orientados a mantener un adecuado flujo de aire hacia los pulmones. En ciertas ocasiones es necesaria la utilización de algunos dispositivos para este fin, tales como cánulas orofaríngeas, mascarilla facial, bolsa para ventilar e incluso llegar a la intubación orotraqueal. Todo ello será revisado en este capítulo. SUMMARY: Airway management in a non difficult airway is an ability every physician should handle. It requires the knowledge of both anatomical and physiological basic concepts to keep a patent air access to the lungs. Ocassionally, for this goal, the use of certain devices such as oropharyngeal cannulaes, facial masks, ventilation bags and even an orotracheal intubation, is necessary. All of this will be reviewed in this chapter. Palabras clave: Vía aérea no difícil, mascarilla facial, intubación orotraqueal, Keywords: Not difficult airway, face mask, orotracheal intubation

  15. HERESY, 2-D Few-Group Static Eigenvalues Calculation for Thermal Reactor

    International Nuclear Information System (INIS)

    Finch, D.R.

    1965-01-01

    1 - Description of problem or function: HERESY3 solves the two- dimensional, few-group, static reactor eigenvalue problem using the heterogeneous (source-sink or Feinburg-Galanin) formalism. The solution yields the reactor k-effective and absorption reaction rates for each rod normalized to the most absorptive rod in the thermal level. Epithermal fissions are allowed at each resonance level, and lattice-averaged values of thermal utilization, resonance escape probability, thermal and resonance eta values, and the fast fission factor are calculated. Kernels in the calculation are based on age-diffusion theory. Both finite reactor lattices and infinitely repeating reactor super-cells may be calculated. Rod parameters may be calculated by several internal options, and a direct interface is provided to a HAMMER system (NESC Abstract 277) lattice library tape to obtain cell parameters. Criticality searches are provided on thermal utilization, thermal eta, and axial leakage buckling. 2 - Method of solution: Direct power iteration on matrix form of the heterogeneous critical equation is used. 3 - Restrictions on the complexity of the problem: Maxima of - 50 flux/geometry symmetry positions; 20 physically different assemblies; 9 resonance levels; 5000 rod coordinate positions

  16. G2 and G3 reactors design; Description des reacteurs G2 et G3

    Energy Technology Data Exchange (ETDEWEB)

    Herreng,; Ertaud,; Pasquet, [Societe Alsacienne de Constructions Mecaniques (France)

    1958-07-01

    'FRANCE ATOME' Manufacturers Party has been entrusted with the G2 and G3 reactors engineering by the french A.E.C., for the first-five-year french project. Although these reactors are essentially plutonium generators, everyone has been linked with a power station which is supposed to supply with 40 MW, 'Electricite de France' has taken the liability upon itself. The reactor core includes most of G1 reactor parts (central gap excluded): horizontal channels, graphite parallelepipedic bricks stacking, steel thermal shield. The cooling is provided with CO{sub 2} under a 15 atmospheres pressure. This pressure is kept steady in a press-stressed concrete packing-case which is a cylinder horizontally shaped. Steel strips tightened encircle the concrete cylinder; itself protected by sole-plates. The cylinder bottom has brought about unusual problems which have been solved by the choice of an hemispheric shape. Packing-case tightness is provided by a 30 mm iron-plate connected with the inner wall of concrete. One of the reactor's special characteristics is the possibility of loading and unloading while operating. On loading side, barrel locks, each weighting 50 tons, allow new cans, at a pressure of 15 atmospheres, to pass. The cans process almost in a steady way through the channel, and finally drop down through bent spouts, then through spiral toboggans into a new lock. The cooling CO{sub 2} flow is provided with 3 turbo-bellows, these are actuated by average pressure-steam, obtained from exchangers. Every reactor supplies 4 exchangers which have been very difficult to build and to set up. The secondary cycle is standard and contains 3 stages (pressure 10,3: 2 and 0,5 kg/cm{sup 2}). Steam can be condensed in the event of a group turbo-generator stopping, with no modifion for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO{sub 2}, its storage and drain. 49 boron carbide rods are used to control the

  17. Investigation of hydrogen-burn damage in the Three Mile Island Unit 2 reactor building

    International Nuclear Information System (INIS)

    Alvares, N.J.; Beason, D.G.; Eidem, G.R.

    1982-06-01

    About 10 hours after the March 28, 1979 Loss-of-Coolant Accident began at Three Mile Island Unit 2, a hydrogen deflagration of undetermined extent occurred inside the reactor building. Examinations of photographic evidence, available from the first fifteen entries into the reactor building, yielded preliminary data on the possible extent and range of hydrogen burn damage. These data, although sparse, contributed to development of a possible damage path and to an estimate of the extent of damage to susceptible reactor building items. Further information gathered from analysis of additional photographs and samples can provide the means for estimating hydrogen source and production rate data crucial to developing a complete understanding of the TMI-2 hydrogen deflagration. 34 figures

  18. A review of the probabilistic safety assessment application to the TR-2 research reactor

    International Nuclear Information System (INIS)

    Goektepe, G.; Adalioglu, U.; Anac, H.; Sevdik, B.; Menteseoglu, S.

    2001-01-01

    A review of the Probabilistic Safety Assessment (PSA) to the TR-2 Research Reactor is presented. The level 1 PSA application involved: selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development, dependent failure analysis. Each of the steps of the analysis given above is reviewed briefly with highlights from the selected results. PSA application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. Insights gained from the application of PSA methodology to the TR-2 research reactor led to a significant safety review of the system

  19. Solid-state track recorder neutron dosimetry in the Three-Mile Island Unit-2 reactor cavity

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.

    1985-04-01

    Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that there are at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity

  20. Neutron dosimetry in the Three-Mile Island Unit 2 reactor cavity with solid-state track recorders

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.; Rao, S.V.; Greenborg, J.; Fricke, V.R.

    1986-01-01

    Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI-2) reactor cavity, for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. This existence of significant neutron streaming also explains the high count rate observed with the source range monitors that are located in the TMI-2 reactor cavity. (author)

  1. Degradation of gas-phase trichloroethylene over thin-film TiO2 photocatalyst in multi-modules reactor

    International Nuclear Information System (INIS)

    Kim, Sang Bum; Lee, Jun Yub; Kim, Gyung Soo; Hong, Sung Chang

    2009-01-01

    The present paper examined the photocatalytic degradation (PCD) of gas-phase trichloroethylene (TCE) over thin-film TiO 2 . A large-scale treatment of TCE was carried out using scale-up continuous flow photo-reactor in which nine reactors were arranged in parallel and series. The parallel or serial arrangement is a significant factor to determine the special arrangement of whole reactor module as well as to compact the multi-modules in a continuous flow reactor. The conversion of TCE according to the space time was nearly same for parallel and serial connection of the reactors.

  2. Bolsa Família e voto na eleição presidencial de 2006: em busca do elo perdido

    Directory of Open Access Journals (Sweden)

    Elaine Cristina Licio

    2009-06-01

    Full Text Available O presente artigo analisa o impacto de ser beneficiário do Programa Bolsa Família do governo federal na decisão de voto na eleição de 2006 e na avaliação atual do Presidente Lula da Silva e contribui para a crescente literatura que explora o impacto desse programa na distribuição de voto em Lula. Contudo, diferentemente de outros estudos, são analisados aqui dados ao nível individual, testando um modelo estatístico multivariado em uma amostra probabilística nacional usando o Barômetro das Américas de 2008. Os resultados indicam um forte impacto de ser beneficiário do programa no voto em Lula e em avaliações positivas de seu desempenho.This article explores the impact of being a Family Grant Program beneficiary in vote choice for President in the 2006 elections and in Lula da Silva's government evaluations. Therefore, the article contributes to the growing literature on how social programs affect voting behaviour in Brazil. However, differently from all other studies, we use individual level data from the AmericasBarometer 2008 Brazilian round, and multivariate statistical analysis to test our hypotheses. Results indicate that being a recipient of the Family Grant Program positively affects vote for Lula and his administration's evaluations.

  3. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  4. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  5. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

  6. A gas-phase reactor powered by solar energy and ethanol for H2 production

    International Nuclear Information System (INIS)

    Ampelli, Claudio; Genovese, Chiara; Passalacqua, Rosalba; Perathoner, Siglinda; Centi, Gabriele

    2014-01-01

    In the view of H 2 as the future energy vector, we presented here the development of a homemade photo-reactor working in gas phase and easily interfacing with fuel cell devices, for H 2 production by ethanol dehydrogenation. The process generates acetaldehyde as the main co-product, which is more economically advantageous with respect to the low valuable CO 2 produced in the alternative pathway of ethanol photoreforming. The materials adopted as photocatalysts are based on TiO 2 substrates but properly modified with noble (Au) and not-noble (Cu) metals to enhance light harvesting in the visible region. The samples were characterized by BET surface area analysis, Transmission Electron Microscopy (TEM) and UV–visible Diffusive Reflectance Spectroscopy, and finally tested in our homemade photo-reactor by simulated solar irradiation. We discussed about the benefits of operating in gas phase with respect to a conventional slurry photo-reactor (minimization of scattering phenomena, no metal leaching, easy product recovery, etc.). Results showed that high H 2 productivity can be obtained in gas phase conditions, also irradiating titania photocatalysts doped with not-noble metals. - Highlights: • A gas-phase photoreactor for H 2 production by ethanol dehydrogenation was developed. • The photocatalytic behaviours of Au and Cu metal-doped TiO 2 thin layers are compared. • Benefits of operating in gas phase with respect to a slurry reactor are presented. • Gas phase conditions and use of not-noble metals are the best economic solution

  7. Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

    International Nuclear Information System (INIS)

    Reuscher, J.A.; Pickard, P.S.

    1976-01-01

    The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

  8. Bolsa Família: insegurança alimentar e nutricional de crianças menores de cinco anos

    Directory of Open Access Journals (Sweden)

    Flávia Monteiro

    2014-05-01

    Full Text Available Estudo transversal descritivo de base populacional realizado no município de Colombo (PR. Os objetivos foram identificar a prevalência de insegurança alimentar das famílias beneficiárias do Programa Bolsa Família e os fatores relacionados a essa condição, bem como descrever o estado nutricional das crianças menores de cinco anos. As análises de associação foram realizadas por meio do teste exato de Fischer. A amostra incluiu 442 famílias, das quais 168 com menores de cinco anos em sua constituição. Para avaliação da insegurança alimentar foi aplicada a Escala Brasileira de Insegurança Alimentar e o estado nutricional das 199 crianças avaliadas foi determinado pelos índices estatura para idade, peso para idade e índice de massa corporal para idade, de acordo com os valores de referência da OMS 2006. A prevalência de insegurança alimentar foi de 81,6%. O excesso de peso e o déficit estatural entre as crianças coexistiram. A insegurança alimentar apresentou-se associada ao índice estatura para idade entre crianças menores de dois anos. A renda familiar per capita e as dívidas alimentares influenciaram significativamente a situação de insegurança alimentar familiar.

  9. Neutron dosimetry in the Three-Mile Island Unit 2 reactor cavity with solid-state track recorders

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.; Rao, S.V.; Greenborg, J.; Fricke, V.R.

    1985-01-01

    Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity

  10. A Conceptual Study of a Supercritical CO2-Cooled Micro Modular Reactor

    Directory of Open Access Journals (Sweden)

    Hwanyeal Yu

    2015-12-01

    Full Text Available A neutronics conceptual study of a supercritical CO2-cooled micro modular reactor (MMR has been performed in this work. The suggested MMR is an extremely compact and truck-transportable nuclear reactor. The thermal power of the MMR is 36.2 MWth and it is designed to have a 20-year lifetime without refueling. A salient feature of the MMR is that all the components including the generator are integrated in a small reactor vessel. For a minimal volume and long lifetime of the MMR core, a fast neutron spectrum is utilized in this work. To enhance neutron economy and maximize the fuel volume fraction in the core, a high-density uranium mono-nitride U15N fuel is used in the fast-spectrum MMR. Unlike the conventional supercritical CO2-cooled fast reactors, a replaceable fixed absorber (RFA is introduced in a unique way to minimize the excess reactivity and the power peaking factor of the core. For a compact core design, the drum-type control absorber is adopted as the primary reactivity control mechanism. In this study, the neutronics analyses and depletions have been performed by using the continuous energy Monte Carlo Serpent code with the evaluated nuclear data file ENDF/B-VII.1 Library. The MMR core is characterized in view of several important safety parameters such as control system worth, fuel temperature coefficient (FTC and coolant void reactivity (CVR, etc. In addition, a preliminary thermal-hydraulic analysis has also been performed for the hottest channel of the Korea Advanced Institute of Science and Technology (KAIST MMR.

  11. Estudios de la biodegradación de cuatro tipos de bolsas oxo - biodegradables empleadas en la venta de productos, utilizando tierra compostable fresca, fresca más aireación y madura, simulando condiciones ambientales de humedad y temperatura del relleno sanitario ubicado en Quito.

    OpenAIRE

    Cadena Calvachi, Daniela Verónica

    2014-01-01

    This work addressed the study of biodegradation of four types of oxo-biodegradable bags simulating the environmental conditions of temperature and humidity Landfill "The Inga" located in the parish Pintag belonging to the province of Pichincha Canton Quito. Este trabajo abordó el estudio de la biodegradación de cuatro tipos de bolsas oxo-biodegradables simulando las condiciones ambientales de temperatura y humedad del Relleno sanitario “El Inga” ubicado en la parroquia Pintag perteneciente...

  12. Analysis of key hardware factors and countermeasure for restricting 49-2 swimming pool reactor lifetime

    International Nuclear Information System (INIS)

    Zhang Yadong; Guo Yue; Yang Xiao; Wang Yiwei; Wang Zhanwen

    2013-01-01

    Safe operation is the most important factor to determine the lifetime of aged 49-2 swimming pool reactor. In this paper, the hardware factors of lifetime were analyzed, such as the pool concrete aging, corrosion of aluminum container and primary coolant system, and graphite swelling etc., and then the corresponding measures such as surveillance, prevention and maintenance were purposed. The results show that 49-2 swimming pool reactor can continue to operate safely due to that container is safe under 8 degree earthquake, the reactor is safe on flood level of once per millennium, adding dam break, and the ageing condition of primary coolant system and container is acceptable. (authors)

  13. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    additional consideration should be required in nuclear design and fuel treating facilities due to reactivity coefficient being shifted to the plus side, larger neutron yield and increased heat source caused by MA loading. (2) Confirmation of TRU burning reactor core concepts. The core specification of sodium cooled-nitride fueled TRU burning large reactor was designed based on commercial type fast reactor (sodium cooled nitride fueled large fast reactor, 38000 MWt) which was designed in the feasibility studies on commercialized fast reactor cycle system. The composition of MAs from LWR's spent fuel was supposed. MA content in the core fuel is settled to 60 wt% based on the JAERI's design in order to maximize the MA transmutation amount. We need to exchange 25% of core fuel with zirconium hydride (ZrH 1.6 ) to attain Doppler coefficient being equivalent to that of the conventional type commercial fast reactor loaded 5 wt% MA. Furthermore, this reactor could transmute MAs produced in forty-eight sodium cooled nitride fueled large fast reactors generating the same output. In order to investigate the dependency of MA transmutation characteristics on the reactor output, 1200 MWt TRU burning middle or small reactor core concept was designed. This core was settled by reducing the number of core fuel assemblies from that of TRU burning large reactor designed above. MA transmutation rate in this core is smaller than that in the TRU burning large reactor core because the neutron flux of this core becomes smaller than that of the TRU burning large reactor core due to the higher Pu enrichment. (3) Comparison between TRU burning reactor and conventional type commercial fast reactor. MA transmutation and nuclear characteristics of the sodium cooled nitride fuel commercial type fast reactor loaded 5 wt%MA were evaluated and compared with those of TRU burning large reactor designed in (2). The commercial type fast reactor could only transmute MAs produced in seven sodium cooled nitride

  14. Properties of an irradiated heat-treated Zr-2.5Nb pressure tube removed from the NPD reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chow, C.K. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Coleman, C.E. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Koike, M.H. [Power Reactor and Nuclear Fuel Development Corp., O-Arai Engineering Centre, O-Arai (Japan); Causey, A.R.; Ells, C.E.; Hosbons, R.R.; Sagat, S.; Urbanic, V.F.; Rodgers, D.K

    1997-07-01

    Some pressure tubes in reactors moderated by heavy water have been made from heat-treated (HT) Zr-2.5Nb. One such tube was removed from the NPD nuclear reactor after 20 years of operation. An extensive program was carried out jointly by AECL and PNC to evaluate the condition and properties of this pressure tube. The investigations include irradiation creep, tensile, corrosion, delayed hydride cracking (DHC), fatigue, and fracture properties. Results show that: (I) the in-reactor elongation rate is much lower and the transverse strain rates are slightly larger than in cold-worked (CW) Zr-2.5Nb tubes; (2) the tensile properties, hydrogen pickup, threshold stress intensity factor for DHC initiation, DHC velocity, and fatigue crack growth rates were similar to those of the CW Zr-2.5Nb material; (3) the fracture toughness of this tube, as measured by curved compact toughness specimens and burst tests, is slightly higher than the CW tubes. The results were also compared with other heat-treated Zr-2.5Nb materials irradiated in the Fugen reactor. The tube was in excellent condition when removed from the reactor and would have been satisfactory for further service. (author)

  15. On-line reactor building integrity testing at Gentilly-2 (summary of results 1987-1994)

    International Nuclear Information System (INIS)

    Collins, N.; Lafreniere, P.

    1994-01-01

    In 1987, Hydro-0uebec embarked on an ambitious development program to provide the Gentilly-2 Nuclear Power Station with an effective and practical Reactor Building Containment integrity Test (CIT). In October 1992, the inaugural low pressure (3 kPa(g) nominal) CIT at 100% F.P was performed. The test was conclusive and the CIT was declared In-Service for containment integrity verification on-line. Five subsequent CITs performed in 1993 and 1994 have demonstrated the expected leak rate results and good reliability. The outstanding feature of the CITs is the demonstrated accurary of better than 5% of the measured leak rate. The CIT was developed with the primary goal of demonstrating 'overall' containment availability. Specifically it was designed to detect a 25 mm. diameter leak or hole in the Reactor Building. However, the remarkable CIT accuracy allows reliable detection of a 2 mm. hole. The Gentilly-2 CIT is an innovative approach based on the Temperature Compensation Method (TCM) which uses a reference volume composed of an extensive tubular network of several different diameters. This eliminates the need to track numerous temperature points. A second independent tubular network includes numerous humidity sampling points, thereby enabling the mearurernent of minute pressure variations inside the Reactor Building, independant of the spatial and temporal humidity behaviour. This Gentilly-2 TOM System has been demonstrated to work at both high and low test pressures. The GentiIly-2 design allows the CIT to be performed at a nominal 3 kPa(g) test pressure during a 12-hour period (28 hours total with alignment time) with the reactor at full power. The traditional Reactor Building Pressure Test (RBPT) is typically performed at high pressure (124 kPa(g) in a 5-day critical path window (7 days total with alignment time) during an annual shutdown

  16. Performance Estimation of Supercritical Co2 Micro Modular Reactor (MMR) for Varying Cooling Air Temperature

    International Nuclear Information System (INIS)

    Ahn, Yoonhan; Kim, Seong Gu; Cho, Seong Kuk; Lee, Jeong Ik

    2015-01-01

    A Small Modular Reactor (SMR) receives interests for the various application such as electricity co-generation, small-scale power generation, seawater desalination, district heating and propulsion. As a part of SMR development, supercritical CO2 Micro Modular Reactor (MMR) of 36.2MWth in power is under development by the KAIST research team. To enhance the mobility, the entire system including the power conversion system is designed for the full modularization. Based on the preliminary design, the thermal efficiency is 31.5% when CO2 is sufficiently cooled to the design temperature. A supercritical CO2 MMR is designed to supply electricity to the remote regions. The ambient temperature of the area can influence the compressor inlet temperature as the reactor is cooled with the atmospheric air. To estimate the S-CO2 cycle performance for various environmental conditions, A quasi-static analysis code is developed. For the off design performance of S-CO2 turbomachineries, the experimental result of Sandia National Lab (SNL) is utilized

  17. Assessment of intestinal permeability and bacterial translocation employing nuclear methods in murine mucositis

    International Nuclear Information System (INIS)

    Pessoa, Rafaela M.; Takenaka, Isabella K.T.M.; Barros, Patricia A.V.; Moura, Livia P.; Contarini, Sara M.L.; Amorim, Juliana M.; Castilho, Raquel O.; Leite, Camila M.A.; Cardoso, Valbert N.; Diniz, Simone Odilia F.

    2017-01-01

    Full text: Introduction: Mucositis affects approximately 80% of patients who receive chemotherapy combinations. The lesions are painful, restrict food intake and make patients more susceptible to systemic infections. Some agents and strategies are being studied for controlling mucositis, none of them is used in clinical practice. In Minas Gerais, many studies have addressed the popular use of the plant Arrabidaea chica in the form of tea, to treat intestinal cramps and diarrhea, the main symptoms of mucositis. Objective: To evaluate the potential of Arrabidaea chica extract in the management of the integrity of the intestinal mucosa, using the experimental model of gut mucositis induced by 5-Fluorouracila (5-FU). Methods: The UFMG Ethics Committee for Animal Experimentation (CETEA/UFMG) approved this study (nº 411/2015). Male BALB/c mice between 6-8 weeks of age were randomly divided into four groups (n=9) as follows: 1. Control (CTL) - oral administration of saline solution (10 days); 2. A. chica (AC) - oral administration of A. chica extract (10 days); 3. Mucositis (MUC) - underwent mucositis (5-FU) (10 days); 4. Mucositis + A. chica (MUC+ AC) - underwent mucositis and received oral administration of A. chica extract (10 days). At the 7 th day, mice in the MUC and MUC + AC groups received an intraperitoneal (IP) injection containing 300 mg/kg 5-FU, whereas the animals of the CTL and AC groups received a saline IP injection. After 72 hours (10 th experimental day), intestinal permeability was determined by measuring the radioactivity diffusion in the blood after oral administration of diethylenetriaminepentaacetic acid (DTPA) labelled with technetium-99m ( 99m Tc) and bacterial translocation was determined by measuring the radioactivity diffusion in the blood after oral administration of E. coli labelled with technetium-99m ( 99m Tc). After 4 hours, the mice were euthanized and assessed for intestinal permeability, bacterial translocation and intestinal histology

  18. Assessment of intestinal permeability and bacterial translocation employing nuclear methods in murine mucositis

    Energy Technology Data Exchange (ETDEWEB)

    Pessoa, Rafaela M.; Takenaka, Isabella K.T.M.; Barros, Patricia A.V.; Moura, Livia P.; Contarini, Sara M.L.; Amorim, Juliana M.; Castilho, Raquel O.; Leite, Camila M.A.; Cardoso, Valbert N.; Diniz, Simone Odilia F. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, Mg (Brazil)

    2017-07-01

    Full text: Introduction: Mucositis affects approximately 80% of patients who receive chemotherapy combinations. The lesions are painful, restrict food intake and make patients more susceptible to systemic infections. Some agents and strategies are being studied for controlling mucositis, none of them is used in clinical practice. In Minas Gerais, many studies have addressed the popular use of the plant Arrabidaea chica in the form of tea, to treat intestinal cramps and diarrhea, the main symptoms of mucositis. Objective: To evaluate the potential of Arrabidaea chica extract in the management of the integrity of the intestinal mucosa, using the experimental model of gut mucositis induced by 5-Fluorouracila (5-FU). Methods: The UFMG Ethics Committee for Animal Experimentation (CETEA/UFMG) approved this study (nº 411/2015). Male BALB/c mice between 6-8 weeks of age were randomly divided into four groups (n=9) as follows: 1. Control (CTL) - oral administration of saline solution (10 days); 2. A. chica (AC) - oral administration of A. chica extract (10 days); 3. Mucositis (MUC) - underwent mucositis (5-FU) (10 days); 4. Mucositis + A. chica (MUC+ AC) - underwent mucositis and received oral administration of A. chica extract (10 days). At the 7{sup th} day, mice in the MUC and MUC + AC groups received an intraperitoneal (IP) injection containing 300 mg/kg 5-FU, whereas the animals of the CTL and AC groups received a saline IP injection. After 72 hours (10{sup th} experimental day), intestinal permeability was determined by measuring the radioactivity diffusion in the blood after oral administration of diethylenetriaminepentaacetic acid (DTPA) labelled with technetium-99m ({sup 99m}Tc) and bacterial translocation was determined by measuring the radioactivity diffusion in the blood after oral administration of E. coli labelled with technetium-99m ({sup 99m}Tc). After 4 hours, the mice were euthanized and assessed for intestinal permeability, bacterial translocation and

  19. VENUS-2, Reactor Kinetics with Feedback, 2-D LMFBR Disassembly Excursions

    International Nuclear Information System (INIS)

    Jackson, J.F.; Nicholson, R.B.; Weber, D.P.

    1980-01-01

    1 - Description of problem or function: VENUS-2 is an improved edition of the VENUS fast-reactor disassembly program. It is a two- dimensional (r-z) coupled neutronics-hydrodynamics code that calculates the dynamic behavior of an LMFBR during a prompt-critical disassembly excursion. It calculates the power history and fission energy release as well as the space-time histories of the fuel temperatures, core material pressures, and core material motions. Reactivity feedback effects due to Doppler broadening and reactor material motion are taken into account. 2 - Method of solution: The power and energy release are calculated using a point-kinetics formulation with up to six delayed neutron groups. The reactivity is a combination of an input driving function and feedback effects due to Doppler broadening and material motion. An adiabatic model is used to calculate the temperature increase throughout the reactor based on an initial temperature distribution and power profile provided as input data. These temperatures are, in turn, converted to fuel pressures through one of several equation of state options provided. The material motion that results from the pressure buildup is calculated by a direct finite difference solution of a set of two-dimensional (r-z) hydrodynamics equations. This is done in Lagrangian coordinates. The reactivity change associated with this motion is calculated by first-order perturbation theory. The displacements are also used to adjust the fuel densities as required for the density dependent equation-of- state option. An automatic time-step-size selection scheme is provided. 3 - Restrictions on the complexity of the problem: VENUS-2 is written so that the dimensions of the storage arrays can be readily changed to accommodate a broad range of problem sizes. In the base version, the total number of mesh intervals is restricted such that (NR+3)*(NZ+3) is less than 700, where NR and NZ are the total number of mesh intervals in the r and z

  20. Energy Multiplier Module (EM{sup 2}) - advanced small modular reactor for electricity generation

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, T.; Schleicher, R.; Choi, H.; Rawls, J., E-mail: timothy.bertch@ga.com [General Atomics, San Diego, California (United States)

    2013-07-01

    In order to provide cost effective nuclear energy in other than large reactor, large grid applications, fission technology needs to make further advances. 'Convert and burn' fast reactors offer long life cores, improved fuel utilization, reduced waste and other benefits while achieving cost effective energy production in a smaller reactor. General Atomics' Energy Multiplier Module (EM{sup 2}), a helium-cooled compact fast reactor that augments its fissile fuel load with either depleted uranium (DU) or used nuclear fuel (UNF). The convert and burn in-situ provides 250 MWe with a 30 year core life. High temperature provides a simple, high efficiency direct cycle gas turbine which along with modular construction, fewer systems, road shipment and minimum on site construction support cost effectiveness. Additional advantages in fuel cycle, non-proliferation and siting flexibility and its ability to meet all safety requirements make for an attractive power source, especially in remote and small grid regions. (author)

  1. Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [SCK CEN (Belgium); Kalcheva, S. [SCK CEN (Belgium); Sikik, E. [SCK CEN (Belgium); Koonen, E. [SCK CEN (Belgium)

    2015-12-01

    BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.

  2. Proceedings of the international topical meeting on advanced reactors safety: Volume 2

    International Nuclear Information System (INIS)

    1997-01-01

    In this volume, 89 papers are grouped under the following headings: advances in research/test reactor safety; advanced reactor accident management and emergency actions; advanced reactors instrumentation/controls/human factors; probabilistic risk/safety and reliability assessments; steam explosion research and issues; advanced reactor severe accident issues and research (analysis and assessments); advanced reactor thermal hydraulics; accelerator-driven source safety; liquid-metal reactor safety; structural assessments and issues; late papers

  3. SEXUALLY TRANSMITTED DISEASES: KNOWLEDGE AND SEXUAL BEHAVIOR OF ADOLESCENTS

    Directory of Open Access Journals (Sweden)

    Niviane Genz

    2017-01-01

    Full Text Available Objetivo: evaluar el conocimiento y comportamiento sexual de los adolescentes acerca de Enfermedades de Transmisión Sexual. Metodo: estudio descriptivo, observacional, cuantitativo, con muestra de conveniencia con 532 adolescentes entre 10 y 19 años. El cuestionario fue administrado sobre ETS. Para el análisis de los datos se utilizó el programa STATA11.1. El proyecto fue aprobado por el. Resultados: 89,2% de las chicas y el 90,3% de los chicos supieron definir adecuadamente el concepto de ETS; 98,5% de las chicas y 98,9% de los chicos el uso del preservativo es el método más eficaz para la prevención. Sin embargo, el 37,1% de las chicas y el 30,5% de los chicos reportaron el uso de anticonceptivos como método preventivo. Conclusion: es saludable la realización de acciones educativas junto a la escuela sobre temas tales como la sexualidad y la salud reproductiva.

  4. Reactor noise analysis of experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

    As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

  5. Application of 2DOF controller for reactor power control. Verification by numerical simulation

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Suzuki, Katsuo

    1996-09-01

    In this report the usefulness of the two degree of freedom (2DOF) control is discussed to improve the reference response characteristics and robustness for reactor power control system. The 2DOF controller consists of feedforward and feedback elements. The feedforward element was designed by model matching method and the feedback element by solving the mixed sensitivity problem of H ∞ control. The 2DOF control gives good performance in both reference response and robustness to disturbance and plant perturbation. The simulation of reactor power control was performed by digitizing the 2DOF controller with the digital control periods of 10[msec]. It is found that the control period of 10[msec] is enough not to make degradation of the control performance by digitizing. (author)

  6. Validation of SCALE4.4a for Calculation of Xe-Sm Transients After a Scram of the BR2 Reactor

    International Nuclear Information System (INIS)

    Kalcheva, S.; Ponsard, B.; Koonen, E.

    2007-01-01

    The aim of this report is to validate the computational modules system SCALE4.4a for evaluation of reactivity changes, macroscopic absorption cross sections and calculations of the positions of the Control Rods during their motion in Xe-Sm transient after a scram of the BR-2 reactor. The rapid shutting down of the reactor by inserting of negative reactivity by the Control Rods is known as a reactor scram. Following reactor scram, a large xenon and samarium buildup occur in the reactor, which may appreciably affect the multiplication factor of the core due to enormous neutron absorption. The validation of the calculations of Xe-Sm transients by SCALE4.4a has been performed on the measurements of the positions of the Control Rods during their motion in Xe-Sm transients of the BR-2 reactor and on comparison with the calculations by the standard procedure XESM, developed at the BR-2 reactor. A final conclusion is made that the SCALE4.4a modules system can be used for evaluation of Xe-Sm transients of the BR-2 reactor. The utilization of the code is simple, the computational time takes from few seconds.

  7. Nuclear powerplant standardization: light water reactors. Volume 2. Appendixes

    International Nuclear Information System (INIS)

    1981-06-01

    This volume contains working papers written for OTA to assist in preparation of the report, NUCLEAR POWERPLANT STANDARDIZATION: LIGHT WATER REACTORS. Included in the appendixes are the following: the current state of standardization, an application of the principles of the Naval Reactors Program to commercial reactors; the NRC and standardization, impacts of nuclear powerplant standardization on public health and safety, descriptions of current control room designs and Duke Power's letter, Admiral Rickover's testimony, a history of standardization in the NRC, and details on the impact of standardization on public health and safety

  8. Degradation of gas-phase trichloroethylene over thin-film TiO{sub 2} photocatalyst in multi-modules reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Bum [New and Renewable Energy Team, Environment and Energy Division, Korea Institute of Industrial Technology (Korea, Republic of); Lee, Jun Yub, E-mail: ljy02191@hanafos.com [Power Engineering Research Institute, Korea Power Engineering Company, Inc. (Korea, Republic of); Kim, Gyung Soo [New and Renewable Energy Team, Environment and Energy Division, Korea Institute of Industrial Technology (Korea, Republic of); Hong, Sung Chang [Department of Environmental Engineering, Kyonggi University (Korea, Republic of)

    2009-07-30

    The present paper examined the photocatalytic degradation (PCD) of gas-phase trichloroethylene (TCE) over thin-film TiO{sub 2}. A large-scale treatment of TCE was carried out using scale-up continuous flow photo-reactor in which nine reactors were arranged in parallel and series. The parallel or serial arrangement is a significant factor to determine the special arrangement of whole reactor module as well as to compact the multi-modules in a continuous flow reactor. The conversion of TCE according to the space time was nearly same for parallel and serial connection of the reactors.

  9. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

    Maillard, M.L.

    1960-01-01

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [fr

  10. Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

    2017-05-15

    Highlights: • Developed low enrichment and natural circulation cooled SLIMM-1.2 SMR for generating 10–100 MW{sub th}. • Neutronics analyses estimate operation life and temperature reactivity feedback. • At 100 MW{sub th}, SLIMM-1.2 operates for 6.3 FPY without refueling. • SLIMM-1.2 has relatively low power peaking and maximum UN fuel temperature < 1400 K. - Abstract: The Scalable LIquid Metal cooled small Modular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100 MW{sub th} continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100 MW{sub th}, the fissile production and depletion, the

  11. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  12. Development of the fast reactor group constant set JFS-3-J3.2R based on the JENDL-3.2

    CERN Document Server

    Chiba, G

    2002-01-01

    It is reported that the fast reactor group constant set JFS-3-J3.2 based on the newest evaluated nuclear data library JENDL3.2 has a serious error in the process of applying the weighting function. As the error affects greatly nuclear characteristics, and a corrected version of the reactor constant set, JFS-3-J3.2R, was developed, as well as lumped FP cross sections. The use of JFS-3-J3.2R improves the results of analyses especially on sample Doppler reactivity and reaction rate in the blanket region in comparison with those obtained using the JFS-3-J3.2.

  13. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  14. Consideration of LH2 and LD2 cold neutron sources in heavy water reactor reflector

    International Nuclear Information System (INIS)

    Potapov, I.A.; Serebrov, A.P.

    2001-01-01

    The reactor power, the required CNS dimensions and power of the cryogenic equipment define the CNS type with maximized cold neutron production. Cold neutron fluxes from liquid hydrogen (LH 2 ) and liquid deuterium (LD 2 ) cold neutron sources (CNS) are analyzed. Different CNS volumes, presents and absence of reentrant holes inside the CNS, different adjustment of beam tube and containment are considered. (orig.)

  15. Fusion reactor materials program plan. Section 2. Damage analysis and fundamental studies

    International Nuclear Information System (INIS)

    1978-07-01

    The scope of this program includes: (1) Development of procedures for characterizing neutron environments of test facilities and fusion reactors, (2) Theoretical and experimental investigations of the influence of irradiation environment on damage production, damage microstructure evolution, and mechanical and physical property changes, (3) Identification and, where appropriate, development of essential nuclear and materials data, and (4) Development of a methodology, based on damage mechanisms, for correlating the mechanical behavior of materials exposed to diverse test environments and projecting this behavior to magnetic fusion reactor (MFR) environments. Some major problem areas are addressed

  16. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

  17. Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwangmin, E-mail: kwangmin81@gmail.com [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Go, Byeong-Soo; Sung, Hae-Jin; Park, Hea-chul; Kim, Seokho [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of); Jin, Yoon-Su; Oh, Yunsang [Vector Fields Korea Inc., Pohang 790-834 (Korea, Republic of); Park, Minwon [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Yu, In-Keun, E-mail: yuik@changwon.ac.kr [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of)

    2014-09-15

    Highlights: • The authors designed and fabricated a D-shape coil based toroid-type HTS DC reactor using 2G GdBCO HTS wires. • The toroid-type magnet consisted of 30 D-shape double pancake coil (DDC)s. The total length of the wire was 2.32 km. • The conduction cooling method was adopted for reactor magnet cooling. • The maximum cooling temperature of reactor magnet is 5.5 K. • The inductance was 408 mH in the steady-state condition (300 A operating). - Abstract: This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

  18. IAEA safety standards for research reactors

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

    The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

  19. Containment Loads Analysis for CANDU6 Reactor using CONTAIN 2.0

    International Nuclear Information System (INIS)

    Kim, Tae H.; Yang, Chae Y.

    2013-01-01

    The containment plays an important role to limit the release of radioactive materials to the environment during design basis accidents (DBAs). Therefore, the containment has to maintain its integrity under DBA conditions. Generally, a containment functional DBA evaluation includes calculations of the key containment loads, i. e., pressure and temperature effects associated with a postulated large rupture of the primary or secondary coolant system piping. In this paper, the behavior of containment pressure and temperature was evaluated for loss of coolant accidents (LOCAs) of the Wolsong unit 1 in order to assess the applicability of CONTAIN 2.0 code for the containment loads analysis of the CANDU6 reactor. The containment pressure and temperature of the Wolsong unit 1 were evaluated using the CONTAIN 2.0 code and the results were compared with the CONTEMPT4 code. The peak pressure and temperature calculated by CONTAIN 2.0 agreed well with those of CONTEMPT4 calculation. The overall result of this analysis shows that the CONTAIN 2.0 code can apply to the containment loads analysis for the CANDU6 reactor

  20. Mass transfer of ammonia escape and CO2 absorption in CO2 capture using ammonia solution in bubbling reactor

    International Nuclear Information System (INIS)

    Ma, Shuangchen; Chen, Gongda; Zhu, Sijie; Han, Tingting; Yu, Weijing

    2016-01-01

    Highlights: • Mass transfer coefficient models of ammonia escape were built. • Influences of temperature, inlet CO 2 and ammonia concentration were studied. • Mass transfer coefficients of ammonia escape and CO 2 absorption were obtained. • Studies can provide the basic data as a reference guideline for process application. - Abstract: The mass transfer of CO 2 capture using ammonia solution in the bubbling reactor was studied; according to double film theory, the mass transfer coefficient models and interface area model were built. Through our experiments, the overall volumetric mass transfer coefficients were obtained, while the interface areas in unit volume were estimated. The volumetric mass transfer coefficients of ammonia escaping during the experiment were 1.39 × 10 −5 –4.34 × 10 −5 mol/(m 3 s Pa), and the volumetric mass transfer coefficients of CO 2 absorption were 2.86 × 10 −5 –17.9 × 10 −5 mol/(m 3 s Pa). The estimated interface area of unit volume in the bubbling reactor ranged from 75.19 to 256.41 m 2 /m 3 , making the bubbling reactor a viable choice to obtain higher mass transfer performance than the packed tower or spraying tower.

  1. O PROGRAMA INSTITUCIONAL DE BOLSA DE INICIAÇÃO À DOCÊNCIA, AS ESCOLHAS PROFISSIONAIS E AS CONDIÇÕES DE TRABALHO DOCENTE

    Directory of Open Access Journals (Sweden)

    Natalia Neves Macedo Deimling

    2017-11-01

    Full Text Available RESUMO: Objetivamos neste artigo apresentar uma análise sobre as influências do Programa Institucional de Bolsa de Iniciação à Docência (PIBID nas escolhas profissionais dos alunos bolsistas da licenciatura que dele participam. Trata-se de uma pesquisa de abordagem qualitativa que tem na entrevista semiestruturada o principal instrumento de construção e análise dos dados. As entrevistas analisadas foram realizadas com seis coordenadores, quatro professores colaboradores e quarenta e oito alunos bolsistas de quatro subprojetos do PIBID de uma universidade federal brasileira no ano de 2013. Os resultados mostram que alguns dos bolsistas entrevistados desejam seguir a carreira docente e que o Programa os tem influenciado positivamente nessa escolha. Todavia, relatos apresentados por outros bolsistas demonstram justamente o desestímulo que eles apresentam pela profissão devido à desvalorização da carreira, aos baixos salários e às condições adversas de trabalho que eles observam nas escolas de educação básica por meio de sua participação no Programa.

  2. SoLid: Search for Oscillations with Lithium-6 Detector at the SCK-CEN BR2 reactor

    Science.gov (United States)

    Ban, G.; Beaumont, W.; Buhour, J. M.; Coupé, B.; Cucoanes, A. S.; D'Hondt, J.; Durand, D.; Fallot, M.; Fresneau, S.; Giot, L.; Guillon, B.; Guilloux, G.; Janssen, X.; Kalcheva, S.; Koonen, E.; Labare, M.; Moortgat, C.; Pronost, G.; Raes, L.; Ryckbosch, D.; Ryder, N.; Shitov, Y.; Vacheret, A.; Van Mulders, P.; Van Remortel, N.; Weber, A.; Yermia, F.

    2016-04-01

    Sterile neutrinos have been considered as a possible explanation for the recent reactor and Gallium anomalies arising from reanalysis of reactor flux and calibration data of previous neutrino experiments. A way to test this hypothesis is to look for distortions of the anti-neutrino energy caused by oscillation from active to sterile neutrino at close stand-off (˜ 6- 8m) of a compact reactor core. Due to the low rate of anti-neutrino interactions the main challenge in such measurement is to control the high level of gamma rays and neutron background. The SoLid experiment is a proposal to search for active-to-sterile anti-neutrino oscillation at very short baseline of the SCK•CEN BR2 research reactor. This experiment uses a novel approach to detect anti-neutrino with a highly segmented detector based on Lithium-6. With the combination of high granularity, high neutron-gamma discrimination using 6LiF:ZnS(Ag) and precise localization of the Inverse Beta Decay products, a better experimental sensitivity can be achieved compared to other state-of-the-art technology. This compact system requires minimum passive shielding allowing for very close stand off to the reactor. The experimental set up of the SoLid experiment and the BR2 reactor will be presented. The new principle of neutrino detection and the detector design with expected performance will be described. The expected sensitivity to new oscillations of the SoLid detector as well as the first measurements made with the 8 kg prototype detector deployed at the BR2 reactor in 2013-2014 will be reported.

  3. A novel condensation reactor for efficient CO2 to methanol conversion for storage of renewable electric energy

    NARCIS (Netherlands)

    Bos, Martin Johan; Brilman, Derk Willem Frederik

    2015-01-01

    A novel reactor design for the conversion of CO2 and H2 to methanol is developed. The conversion limitations because of thermodynamic equilibrium are bypassed via in situ condensation of a water/methanol mixture. Two temperatures zones inside the reactor ensure optimal catalyst activity (high

  4. Quality assurance in the project of RECH-2 research reactor

    International Nuclear Information System (INIS)

    Goycolea Donoso, C.; Nino de Zepeda Schele, A.

    1989-01-01

    The implantation of a Quality Assurance Program for the design, supply, construction, installation, and testing of the RECH-2 research reactor, is described in this paper. The obtained results, demonstrate that a Quality Assurance Program constitutes a suitable mean to assure that the installation complies with the safety and reliability requirements. (author)

  5. Set of rules SOR 2 licensing of nuclear reactors

    International Nuclear Information System (INIS)

    1976-05-01

    This is the set of rules promulgated by the Israel Atomic Energy Commission pursuant to the Supervision of Supplies and Services Law 5718-1957, Order regarding Supervision of Nuclear Reactors (1974) Chapter 3: Permits, to provide for the Licensing of Nuclear Reactors. (B.G.)

  6. Measurement of thermal conductivity of sintered UO{sub 2} in the reactor; Merenje toplotne provodljivosti sinterovanog UO{sub 2} u reaktoru

    Energy Technology Data Exchange (ETDEWEB)

    Katanic, J; Stevanovic, M [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

    1965-10-15

    Thermal conductivity is considered one of the fundamental properties of sintered UO{sub 2} fuel. Samples should be tested under real core conditions. This paper covers the methods and instruments for thermal conductivity measurement of UO{sub 2} samples in the reactor core, measurements outside the core under conditions similar to those in the core and outside the core after irradiation. Fuel samples are placed in capsules for irradiation in the reactor in-core loops.

  7. Geomechanical Analysis of Underground Coal Gasification Reactor Cool Down for Subsequent CO2 Storage

    Science.gov (United States)

    Sarhosis, Vasilis; Yang, Dongmin; Kempka, Thomas; Sheng, Yong

    2013-04-01

    Underground coal gasification (UCG) is an efficient method for the conversion of conventionally unmineable coal resources into energy and feedstock. If the UCG process is combined with the subsequent storage of process CO2 in the former UCG reactors, a near-zero carbon emission energy source can be realised. This study aims to present the development of a computational model to simulate the cooling process of UCG reactors in abandonment to decrease the initial high temperature of more than 400 °C to a level where extensive CO2 volume expansion due to temperature changes can be significantly reduced during the time of CO2 injection. Furthermore, we predict the cool down temperature conditions with and without water flushing. A state of the art coupled thermal-mechanical model was developed using the finite element software ABAQUS to predict the cavity growth and the resulting surface subsidence. In addition, the multi-physics computational software COMSOL was employed to simulate the cavity cool down process which is of uttermost relevance for CO2 storage in the former UCG reactors. For that purpose, we simulated fluid flow, thermal conduction as well as thermal convection processes between fluid (water and CO2) and solid represented by coal and surrounding rocks. Material properties for rocks and coal were obtained from extant literature sources and geomechanical testings which were carried out on samples derived from a prospective demonstration site in Bulgaria. The analysis of results showed that the numerical models developed allowed for the determination of the UCG reactor growth, roof spalling, surface subsidence and heat propagation during the UCG process and the subsequent CO2 storage. It is anticipated that the results of this study can support optimisation of the preparation procedure for CO2 storage in former UCG reactors. The proposed scheme was discussed so far, but not validated by a coupled numerical analysis and if proved to be applicable it could

  8. Análisis de portafolio por sectores mediante el uso de algoritmos genéticos: caso aplicado a la Bolsa Mexicana de Valores

    Directory of Open Access Journals (Sweden)

    Martha del Pilar Rodríguez García

    2015-01-01

    Full Text Available El tipo de sector, el tamaño de la empresa, el número de trabajadores, etc. son variables que se consideran de control en una gran cantidad de publicaciones. En este trabajo consideramos estudiar la variable sector —más que como una variable de control— como una variable determinante del desempeño financiero (Baird et al. 2012 y del riesgo (Artikis y Nifora, 2011. Así, se analiza seis sectores de la economía mexicana divididos de acuerdo con la Bolsa Mexicana de Valores en Industrial, Productos de consumo básico, Materiales, Productos de consumo no básico, Telecomunicaciones y Servicios financieros. La muestra se compone de 30 empresas mexicanas dentro del periodo de 2007-2012. Para medir el desempeño del portafolio se utilizan dos indicadores clásicos: (1 Alfa de Jensen y (2 Ratio de Sharpe; se utiliza una métrica condicional que mide el número de veces que el rendimiento del portafolio supera el rendimiento promedio del mercado. El objetivo es encontrar un portafolio que maximice estos parámetros y comparar los resultados entre los diferentes sectores bajo estudio. Debido a un problema de programación no lineal, se utilizan algoritmos genéticos para obtener el portafolio óptimo que maximice estas métricas. Los resultados muestran un mejor desempeño financiero ajustado a riesgo en el sector de Materiales y Servicios financieros y un desempeño más bajo en sectores como el Industrial y el de Telecomunicaciones.

  9. Analysis of SBO accident and natural circulation of 49-2 swimming pool reactor

    International Nuclear Information System (INIS)

    Wu Yuanyuan; Liu Tiancai; Sun Wei

    2012-01-01

    The transient thermal hydraulic characteristics of 49-2 Swimming Pool Reactor (SPR) were analyzed by RELAP5/MOD3.3 code to verify the capability of natural circulation and minus reactivity feedback for accident mitigation under the condition of station blackout (SBO). Then, the effects on accident consequence and sequence for core channels and primary pumps were briefly discussed. The calculation results show that the reactor can be shutdown by the effect of minus reactivity feedback, and the residual heat can be removed through the stable natural circulation. Therefore, it demonstrates that the 49-2 SPR is safe during the accident of SBO. (authors)

  10. Status of IVO-FR2-Vg7 experiment for irradiation of fast reactor fuel rods

    International Nuclear Information System (INIS)

    Elbel, H.; Kummerer, K.; Bojarsky, K.; Lopez Jimenez, J.; Otero de la Gandara, J.L.

    1979-01-01

    Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid) concerning a Joint Irradiation Program of Fast Reactor Fuel Rods. The design of fuel rods in general is defined, and, in particular of those with a density 94% DT and diameter 7.6 mm up to a burn-up of 7% FIMA, to be irradiated in the FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used in this irradiation, other possibilities of irradiation in the reactor will also be described. (auth.)

  11. CANDU reactors with reactor grade plutonium/thorium carbide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

    2011-08-15

    Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

  12. Revision of fast reactor group constant set JFS-3-J2

    International Nuclear Information System (INIS)

    Takano, Hideki; Kaneko, Kunio.

    1989-10-01

    To improve the fast reactor group constant set JFS-3-J2 to be applicable for high burnup reactor calculations, group constants for 155 fission product nuclides and the lumped group cross sections for four mother fission isotopes of U-235, U-238, Pu-239 and Pu-241 have been generated. Furthermore, the group constants for higher actinides such as Am and Cm have been produced on the basis of the JENDL-2 nuclear data, so as to be able to use for TRU-transmutation calculations. Benchmark test of this revised set has been performed by analysing the 21 fast critical experimental assemblies. Benchmark calculation system based on one-dimensional Sn-method has been developed to investigate the accuracy of one-dimensional diffusion calculations. Significant difference between the results obtained with the diffusion and transport calculations was observed for small cores and the assemblies with iron or nickel reflector. (author)

  13. UCLA program in reactor studies: The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

  14. N2O Catalytic Decomposition – from Laboratory Experiment to Industry Reactor

    Czech Academy of Sciences Publication Activity Database

    Obalová, L.; Jirátová, Květa; Karásková, K.; Chromčáková, Ž.

    2012-01-01

    Roč. 191, č. 1 (2012), s. 116-120 ISSN 0920-5861 R&D Projects: GA TA ČR TA01020336 Institutional support: RVO:67985858 Keywords : N2O * catalytic decomposition * fixed bed reactor Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 2.980, year: 2012

  15. Nuclear Reactor RA Safety Report, Vol. 4, Reactor

    International Nuclear Information System (INIS)

    1986-11-01

    RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

  16. Alteration of installation of reactors (alteration of No.1 and No.2 reactor facilities) in Oi Power Station, Kansai Electric Power Co., Inc

    International Nuclear Information System (INIS)

    1984-01-01

    The Nuclear Safety Commission reported to the Minister of International Trade and Industry on October 27, 1983, that the technical capability was recognized to be adequate, and the safety after the alteration of the installation of reactors was judged to be ensured. At the time of deliberation, the guidelines for examining the safety design and safety evaluation of LWR facilities for power generation were used. Regarding the change of the degree of enrichment of replacement fuel from 3.2 to 3.4 wt.%, the limiting conditions are satisfied in the replacement core, and the nuclear design is appropriate. Eight test fuel assemblies using UO 2 pellets containing gadolinia are charged in the core of No.2 reactor, and the irradiation of two cycles is carried out. As the result of the safety examination regarding this test, the propriety of the nuclear design and mechanical design of the test fuel assemblies was confirmed. This alteration does not exert influence on the result of safety analysis made so far. This report was decided by the Committee on Examination of Reactor Safety based on the conclusion of No.26 subcommittee. (Kako, I.)

  17. Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

    International Nuclear Information System (INIS)

    Ravnik, M.; Mele, I.; Trkov, A.; Rant, J.; Glumac, B.; Dimic, V.

    2008-01-01

    TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2

  18. Plasma-catalyst hybrid reactor with CeO2/γ-Al2O3 for benzene decomposition with synergetic effect and nano particle by-product reduction.

    Science.gov (United States)

    Mao, Lingai; Chen, Zhizong; Wu, Xinyue; Tang, Xiujuan; Yao, Shuiliang; Zhang, Xuming; Jiang, Boqiong; Han, Jingyi; Wu, Zuliang; Lu, Hao; Nozaki, Tomohiro

    2018-04-05

    A dielectric barrier discharge (DBD) catalyst hybrid reactor with CeO 2 /γ-Al 2 O 3 catalyst balls was investigated for benzene decomposition at atmospheric pressure and 30 °C. At an energy density of 37-40 J/L, benzene decomposition was as high as 92.5% when using the hybrid reactor with 5.0wt%CeO 2 /γ-Al 2 O 3 ; while it was 10%-20% when using a normal DBD reactor without a catalyst. Benzene decomposition using the hybrid reactor was almost the same as that using an O 3 catalyst reactor with the same CeO 2 /γ-Al 2 O 3 catalyst, indicating that O 3 plays a key role in the benzene decomposition. Fourier transform infrared spectroscopy analysis showed that O 3 adsorption on CeO 2 /γ-Al 2 O 3 promotes the production of adsorbed O 2 - and O 2 2‒ , which contribute benzene decomposition over heterogeneous catalysts. Nano particles as by-products (phenol and 1,4-benzoquinone) from benzene decomposition can be significantly reduced using the CeO 2 /γ-Al 2 O 3 catalyst. H 2 O inhibits benzene decomposition; however, it improves CO 2 selectivity. The deactivated CeO 2 /γ-Al 2 O 3 catalyst can be regenerated by performing discharges at 100 °C and 192-204 J/L. The decomposition mechanism of benzene over CeO 2 /γ-Al 2 O 3 catalyst was proposed. Copyright © 2017 Elsevier B.V. All rights reserved.

  19. COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors

    International Nuclear Information System (INIS)

    Kaminaga, Masanori

    1994-03-01

    The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode as well as COOLOD-N code. In the COOLOD-N2 code, a 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. (author)

  20. Disputas, ajustes e acomodações na produção da agenda eleitoral: a cobertura jornalística ao Programa Bolsa Família e as eleições de 2006

    Directory of Open Access Journals (Sweden)

    Flávia Biroli

    2010-06-01

    Full Text Available Resumo: Este artigo apresenta uma análise da cobertura jornalística ao Programa Bolsa Família durante as eleições presidenciais de 2006. A pesquisa abrange um total de 166 textos que mencionaram o Programa, publicados pelos jornais O Globo, O Estado de São Paulo, Folha de São Paulo e Valor Econômico entre os dias 1º de setembro e 31 de outubro de 2006. Discutimos a dinâmica de produção da agenda eleitoral, observando as disputas, ajustes e acomodações que constituem a cobertura. A análise conjunta das vozes e dos enquadramentos presentes no material permite observar aspectos relevantes das interações entre os campos da mídia e da política no contexto em que a cobertura foi realizada. As conclusões ressaltam a baixa pluralidade do noticiário, associada a representações das eleições de 2006 e da democracia brasileira que têm como aspectos centrais a estigmatização dos eleitores de baixa renda e dos beneficiários de programas sociais.Abstract: This article presents an analysis of the news about an important social program maintained by the federal government, Programa Bolsa Família, in the period of brazilian major elections of 2006. The study is based in 166 texts published in the newspapers O Globo, O Estado de São Paulo, Folha de São Paulo e Valor Econômico between September 1st and October 31st. We discuss the production of electoral agenda, observing disputes, adjustments and acommodations that constitute news coverage. The analysis of voices and framings in the texts leads to the observation of relevant aspects of the relations between media and politics at that moment. Conclusions underline the low plurality in press coverage, connected to representations of the elections of 2006 and Brazilian democracy that include stigmatization of low income voters and beneficiaries of social programs.

  1. Ethanol production by immobilized yeast and its CO2 gas effects on a packed bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, G M; Choi, C Y; Choi, Y D; Han, M H

    1982-10-01

    Immobilised yeast trapped in an alginate matrix demonstrated maximum activity at 30 degrees C and showed no pH effect between 3 and 7. Substrate inhibition was observed at glucose concentrations above 8% but the immobilised cells retained 70% of their maximum activity at 20% glucose concentration. The operation stability of immobilised cells was lower in simple glucose solution than in the activation medium in which only 20% of the activity was lost after 10 days operation. Inactivated immobilised yeast beads were reactivated by incubation in activation medium without a significant increase in cell numbers in a bead. During the operation of the immobilised yeast in a packed bed reactor, CO/sub 2/ gas accumulation adversely affected the reactor performance. An ideal plus flow reactor, not taking into account the formation of CO/sub 2/ gas bubbles and the presence of mass trasnfer resistance, was simulated using a kinetic model for the production of ethanol and the simulation results were compared with the actual reactor performance to determine the CO/sub 2/ gas effect, quantitatively. Up to 45% of the substrate conversion was lost due to the accumulation of CO/sub 2/ gas bubbles in all cases. (Refs. 21).

  2. Power Quality Problems Mitigation using Dynamic Voltage Restorer in Egypt Thermal Research Reactor (ETRR-2)

    International Nuclear Information System (INIS)

    Kandil, T.; Ayad, N.M.; Abdel Haleam, A.; Mahmoud, M.

    2013-01-01

    Egypt thermal research reactor (ETRR-2) was subjected to several Power Quality Problems such as voltage sags/swells, harmonics distortion, and short interruption. ETRR-2 encompasses a wide range of loads which are very sensitive to voltage variations and this leads to several unplanned shutdowns of the reactor due to trigger of the Reactor Protection System (RPS). The Dynamic Voltage Restorer (DVR) has recently been introduced to protect sensitive loads from voltage sags and other voltage disturbances. It is considered as one of the most efficient and effective solution. Its appeal includes smaller size and fast dynamic response to the disturbance. This paper describes a proposal of a DVR to improve power quality in ETRR-2 electrical distribution systems . The control of the compensation voltage is based on d-q-o algorithm. Simulation is carried out by Matlab/Simulink to verify the performance of the proposed method

  3. Modernization of turbine control system and reactor control system in Almaraz 1 and 2; MOdernizacion de los sistemas de control de turbina y del reactor en Almaraz 1 y 2

    Energy Technology Data Exchange (ETDEWEB)

    Pulido, C.; Diez, J.; Carrasco, J. A.; Lopez, L.

    2005-07-01

    The replacement of the Turbine Control System and Reactor Control System are part of the Almaraz modernization program for the Instrumentation and Control. For these upgrades Almaraz has selected the Ovation Platform that provides open architecture and easy expansion to other systems, these platforms is highly used in many nuclear and thermal plants around the world. One of the main objective for this project were to minimize the impact on the installation and operation of the plant, for that reason the project is implemented in two phases, Turbine Control upgrade and Reactor Control upgrade. Another important objective was to increase the reliability of the control system making them fully fault tolerant to single failures. The turbine Control System has been installed in Units 1 and 2 while the Reactor Control System will be installed in 2006 and 2007 outages. (Author)

  4. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  5. Study of the obtainment of Mo_2C by gas-solid reaction in a fixed and rotary bed reactor

    International Nuclear Information System (INIS)

    Araujo, C.P.B. de; Souza, C.P. de; Souto, M.V.M.; Barbosa, C.M.; Frota, A.V.V.M.

    2016-01-01

    Carbides' synthesis via gas-solid reaction overcomes many of the difficulties found in other processes, requiring lower temperatures and reaction times than traditional metallurgic routes, for example. In carbides' synthesis in fixed bed reactors (FB) the solid precursor is permeated by the reducing/carburizing gas stream forming a packed bed without mobility. The use of a rotary kiln reactor (RK) adds a mixing character to this process, changing its fluid-particle dynamics. In this work ammonium molybdate was subjected to carbo-reduction reaction (CH4 / H2) in both reactors under the same gas flow (15L / h) and temperature (660 ° C) for 180 minutes. Complete conversion was observed Mo2C (dp = 18.9nm modal particles sizes' distribution) in the fixed bed reactor. In the RK reactor this conversion was only partial (∼ 40%) and Mo2C and MoO3 (34nm dp = bimodal) could be observed on the produced XRD pattern. Partial conversion was attributed to the need to use higher solids loading in the reactor CR (50% higher) to avoid solids to centrifuge. (author)

  6. 2-DB, 2-D Multigroup Diffusion, X-Y, R-Theta, Hexagonal Geometry Fast Reactor, Criticality Search

    International Nuclear Information System (INIS)

    Little, W.W. Jr.; Hardie, R.W.; Hirons, T.J.; O'Dell, R.D.

    1969-01-01

    1 - Description of problem or function: 2DB is a flexible, two- dimensional (x-y, r-z, r-theta, hex geometry) diffusion code for use in fast reactor analyses. The code can be used to: (a) Compute fuel burnup using a flexible material shuffling scheme. (b) Perform criticality searches on time absorption (alpha), material concentrations, and region dimensions using a regular or adjoint model. Criticality searches can be performed during burnup to compensate for fuel depletion. (c) Compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Standard source-iteration techniques are used. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy (group) averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes, are formed by the user. The code does not contain built-in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated. The current 1108 version, however, is nominally restricted to 50 energy groups in a 65 K memory. In the 6600 version the power fraction, average burnup rate, and breeding ratio calculations are limited to reactors with a maximum of 50 zones

  7. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Memmott, Matthew [Brigham Young Univ., Provo, UT (United States); Boy, Guy [Florida Inst. of Technology, Melbourne, FL (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Lee, John [Univ. of Michigan, Ann Arbor, MI (United States); Muldrow, Lycurgus [Morehouse College, Atlanta, GA (United States); Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, Wesley [Univ. of Tennessee, Knoxville, TN (United States); Haghighat, Alierza [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States)

    2017-10-02

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I2S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

  8. Operating reactors licensing actions summary. Vol. 4, No. 2

    International Nuclear Information System (INIS)

    1984-04-01

    This summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

  9. Experimental estimations of the kinetics parameters of the IBR-2M reactor by stochastic noises

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Tajybov, L.A.; Garibov, A.A.; Mekhtieva, R.N.

    2012-01-01

    Experimental investigations of stochastic fluctuations of pulse energy of the IBR-2M reactor have been carried out which allowed us to obtain some of the parameters of the reactor kinetics. At different levels of average power a sequence of values of pulse energy was recorded with the calculation of the distribution parameters. An ionization chamber with boron installed near the active zone was used as a neutron detector. The research results allowed us to estimate the average lifetime of prompt neutrons τ = (6.53±0.2)·10 -8 s, absolute power of the reactor and intensity of the source of spontaneous neutrons S sp ≤(6.72±0.12)·10 6 s -1 . It was shown that the experimental results are close to the calculated ones

  10. Selección de una cartera de valores mediante la aplicación de métodos multiobjetivo interactivos a datos reales de la Bolsa española.

    Directory of Open Access Journals (Sweden)

    Mariano Luque Gallego,

    2004-01-01

    Full Text Available En este trabajo aplicamos diversos métodos multiobjetivo interactivos a datos reales de la bolsa española, en concreto datos semanales del periodo 1995-2002. En nuestro modelo consideramos 5 funciones objetivo relacionadas con el deseo del decisor de maximizar la rentabilidad obtenida soportando el menor riesgo posible. Así, tratamos de maximizar la rentabilidad, minimizar la beta de la cartera como representante del riesgo sistemático, minimizar la desviación estándar y la covarianza las cuales recogen el riesgo global soportado y, por último, minimizar la varianza de los residuos como representante del riesgo específico. Tras obtener una solución mediante el método interactivo G-D-F, y tras solicitar información sobre sus preferencias al decisor, vamos cambiando de método para aprovechar las ventajas de cada uno hasta obtener una solución aceptada por el decisor.

  11. Enhancements to the SLOWPOKE-2 nuclear research reactor at the Royal Military College of Canada

    Energy Technology Data Exchange (ETDEWEB)

    Hungler, P.C.; Andrews, M.T.; Weir, R.D.; Nielson, K.S.; Chan, P.K.; Bennett, L.G.I., E-mail: paul.hungler@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

    2014-07-01

    In 1985 a Safe Low Power C(K)ritical Experiment (SLOWPOKE) nuclear research reactor was installed at the Royal Military College of Canada (RMCC). The reactor at nominally 20 kW thermal was named SLOWPOKE-2 and the core was designed to have a total of 198 fuel pins with Low Enriched Uranium (LEU) fuel (19.89% U-235). Installation of the reactor was intended to provide an education tool for members of the Canadian Armed Forces (CAF) and an affordable neutron source for the application of neutron activation analysis (NAA) and radioisotope production. Today, the SLOWPOKE-2 at RMCC continues to be a key education tool for undergraduate and post-graduate students and successfully conducts NAA and isotope production as per its original design intent. RMCC has significantly upgraded the facility and instruments to develop capabilities such as delayed neutron and gamma counting (DNGC) and neutron imaging, including 2D thermal neutron radiography and 3D thermal neutron tomography. These unique nuclear capabilities have been applied to relevant issues in the CAF. The analog control system originally installed in 1985 has been removed and replaced in 2001 by the SLOWPOKE Integrated Reactor Control and Instrumentation System (SIRCIS) which is a digital controller. This control system continues to evolve with SIRCIS V2 currently in operation. The continual enhancement of the facility, instruments and systems at the SLOWPOKE-2 at RMCC will be discussed, including an update on RMCC's refueling plan. (author)

  12. Enhancements to the SLOWPOKE-2 nuclear research reactor at the Royal Military College of Canada

    International Nuclear Information System (INIS)

    Hungler, P.C.; Andrews, M.T.; Weir, R.D.; Nielson, K.S.; Chan, P.K.; Bennett, L.G.I.

    2014-01-01

    In 1985 a Safe Low Power C(K)ritical Experiment (SLOWPOKE) nuclear research reactor was installed at the Royal Military College of Canada (RMCC). The reactor at nominally 20 kW thermal was named SLOWPOKE-2 and the core was designed to have a total of 198 fuel pins with Low Enriched Uranium (LEU) fuel (19.89% U-235). Installation of the reactor was intended to provide an education tool for members of the Canadian Armed Forces (CAF) and an affordable neutron source for the application of neutron activation analysis (NAA) and radioisotope production. Today, the SLOWPOKE-2 at RMCC continues to be a key education tool for undergraduate and post-graduate students and successfully conducts NAA and isotope production as per its original design intent. RMCC has significantly upgraded the facility and instruments to develop capabilities such as delayed neutron and gamma counting (DNGC) and neutron imaging, including 2D thermal neutron radiography and 3D thermal neutron tomography. These unique nuclear capabilities have been applied to relevant issues in the CAF. The analog control system originally installed in 1985 has been removed and replaced in 2001 by the SLOWPOKE Integrated Reactor Control and Instrumentation System (SIRCIS) which is a digital controller. This control system continues to evolve with SIRCIS V2 currently in operation. The continual enhancement of the facility, instruments and systems at the SLOWPOKE-2 at RMCC will be discussed, including an update on RMCC's refueling plan. (author)

  13. Collective occupational dose for nuclear reactors of the 2., 3. and 4. generation

    International Nuclear Information System (INIS)

    Guidez, J.; Saturnin, A.

    2016-01-01

    In France during reactor operation the individual occupational doses are collected and recorded according to the law. When you sum up all the individual doses you get the yearly collective dose expressed in Man.Sv/year. This piece of information can be used to make comparisons between various types of reactors and between reactors of the same type. The results show a steady decrease of the collective dose for all types of reactors over the time except for CANDU reactors for which a slight increase of the dose has appeared since the years 1996-1998. The decrease is due to the continuous improvement of reactor operating and to changes in the reactor design. There is also a constant gap over time between the collective dose for a BWR reactor (1.12 Man.Sv/y) and a PWR reactor 0.60 Man.Sv/y), this gap is certainly due to N 16 nuclide that is created in the primary circuit and transported to turbines in the case of a BWR reactor. For sodium-cooled fast reactors (RNR-Na) the collective dose is below 0.40 Man.Sv/y except for the BN-600 reactor. (A.C.)

  14. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    Veeraraghaven, N.

    1990-01-01

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  15. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    2010-03-01

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  16. Safety of nuclear power reactors

    International Nuclear Information System (INIS)

    MacPherson, H.G.

    1982-01-01

    Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

  17. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-07-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities.

  18. Calibration of RB reactor power

    International Nuclear Information System (INIS)

    Sotic, O.; Markovic, H.; Ninkovic, M.; Strugar, P.; Dimitrijevic, Z.; Takac, S.; Stefanovic, D.; Kocic, A.; Vranic, S.

    1976-09-01

    The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8√2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation

  19. Status of French reactors

    International Nuclear Information System (INIS)

    Ballagny, A.

    1997-01-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm 3 . The OSIRIS reactor has already been converted to LEU. It will use U 3 Si 2 as soon as its present stock of UO 2 fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU

  20. Physical measurements at the RA reactor related to VISA-2, e. Measurements of flux and reactivity during RA reactor operation and exploitation; Fizicka merenja na reaktoru RA u vezi projekta VISA-2, e. Pracenje fluksa i reaktivnosti u toku eksploatacije reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-05-15

    This report includes the following: characteristics of neutron flux in vertical experimental channels of the RA reactor; characteristics of neutron flux in VISA-2 channels; reactivity changes in the reactor during VISA-2 irradiation including calibration of control rods.

  1. Passive Decay Heat Removal System Options for S-CO2 Cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Moon, Jangsik; Jeong, Yong Hoon; Lee, Jeong Ik

    2014-01-01

    To achieve modularization of whole reactor system, Micro Modular Reactor (MMR) which has been being developed in KAIST took S-CO 2 Brayton power cycle. The S-CO 2 power cycle is suitable for SMR due to high cycle efficiency, simple layout, small turbine and small heat exchanger. These characteristics of S-CO 2 power cycle enable modular reactor system and make reduced system size. The reduced size and modular system motived MMR to have mobility by large trailer. Due to minimized on-site construction by modular system, MMR can be deployed in any electricity demand, even in isolated area. To achieve the objective, fully passive safety systems of MMR were designed to have high reliability when any offsite power is unavailable. In this research, the basic concept about MMR and Passive Decay Heat Removal (PDHR) system options for MMR are presented. LOCA, LOFA, LOHS and SBO are considered as DBAs of MMR. To cope with the DBAs, passive decay heat removal system is designed. Water cooled PDHR system shows simple layout, but has CCF with reactor systems and cannot cover all DBAs. On the other hand, air cooled PDHR system with two-phase closed thermosyphon shows high reliability due to minimized CCF and is able to cope with all DBAs. Therefore, the PDHR system of MMR will follows the air-cooled PDHR system and the air cooled system will be explored

  2. A study of UO2 wafer fuel for very high-power research reactors

    International Nuclear Information System (INIS)

    Hsieh, T.C.; Jankus, V.Z.; Rest, J.; Billone, M.C.

    1983-01-01

    The Reduced Enrichment Research and Test Reactor Program is aimed at reducing fuel enrichment to 2 caramel fuel is one of the most promising new types of reduced-enrichment fuel for use in research reactors with very high power density. Parametric studies have been carried out to determine the maximum specific power attainable without significant fission-gas release for UO 2 wafers ranging from 0.75 to 1.50 mm in thickness. The results indicate that (1) all the fuel designs considered in this study are predicted not to fail under full power operation up to a burnup, of 1.9x10 21 fis/cm 3 ; (2) for all fuel designs, failure is predicted at approximately the same fuel centerline temperature for a given burnup; (3) the thinner the wafer, the wider the margin for fuel specific power between normal operation and increased-power operation leading to fuel failure; (4) increasing the coolant pressure in the reactor core could improve fuel performance by maintaining the fuel at a higher power level without failure for a given burnup; and (5) for a given power level, fuel failure will occur earlier at a higher cladding surface temperature and/or under power-cycling conditions. (author)

  3. Microstructure in Zircaloy Creep Tested in the R2 Reactor

    International Nuclear Information System (INIS)

    Pettersson, Kjell

    2004-12-01

    Tubular specimens of Zircaloy-4 have been creep tested in bending in the R2 reactor in Studsvik. The creep deformation in the reactor core is accelerated in comparison with creep deformation outside the reactor core. The possible mechanisms behind this behaviour are described briefly. In order to determine which the actual mechanism is, the microstructure of the material creep tested in the R2 reactor has been examined by transmission electron microscopy. Due to the bending, material subjected to both tensile and compressive stress during creep was available. Since some of the proposed mechanisms might give microstructures which are different when the material is subjected to compressive or tensile stress it was assumed that examination of both types of material would give valuable information with regard to the operating mechanism. The result of the examination was that in the as-irradiated condition there were no obvious differences detected between materials which had been deformed in tension or compression. After a heat treatment to coarsen the irradiation induced microstructure there were still no significant differences between the two types of material. However it was now observed that in addition to dislocation loops the microstructure also contained network dislocations which presumably had been invisible in the electron microscope before heat treatment due to the high density of small dislocation loops in this state. It is therefore concluded that the most probable mechanism for irradiation creep in this case is climb and glide of the network dislocations. The role of irradiation is two-fold: It accelerates climb due to the production of point defects of which more interstitials than vacancies arrive to the network dislocations stopped at an obstacles. This leads to a net climb after which a dislocation is released from the obstacle and an amount of glide takes place. The second effect is the production of loops which serve as an increasing density of

  4. In-situ stripping of H{sub 2}S in gasoil hydrodesulphurization - reactor design considerations

    Energy Technology Data Exchange (ETDEWEB)

    Nava, J.A.O.; Krishna, R. [Amsterdam Univ., Dept. of Chemical Engineering, Amsterdam (Netherlands)

    2004-02-01

    In order to meet future diesel specifications the sulphur content of diesel would need to be reduced to below 50 ppm. This requirement would require improved reactor configurations. In this study we examine the benefits of counter-current contacting of gas oil with H{sub 2}, over conventional co-current contacting in a trickle bed hydrodesulphurization (HDS) reactor. In counter-current contacting, we achieve in-situ stripping of H{sub 2}S from the liquid phase; this is beneficial to the HDS kinetics. A comparison simulation study shows that counter-current contacting would require about 20% lower catalyst load than co-current contacting. However, counter-current contacting of gas and liquid phases in conventionally used HDS catalysts, of 1.5 mm sizes, is not possible due to flooding limitations. The catalysts need to be housed in special wire gauze envelopes as in the catalytic bales or KATAPAK-S configurations. A preliminary hardware design of a counter-current HDS reactor using catalytic bales was carried out in order to determine the technical feasibility. Using a realistic sulphur containing feedstock, the target of 50 ppm S content of desulphurized oil could be met in a reactor of reasonable dimensions. The study also underlines the need for accurate modelling of thermal effects during desulphurization. Our study also shows that interphase mass transfer is unlikely to be a limiting factor and there is a need to develop improved reactor configurations allowing for increased catalyst loading, at the expense of gas-liquid interfacial area. (Author)

  5. Accidents of loss of flow for the ETTR-2 reactor; deterministic analysis

    International Nuclear Information System (INIS)

    El-Messiry, A.M.

    2000-01-01

    The main objective for reactor safety is to keep the fuel in a thermally safe condition with adequate safety margins during all operational modes (normal-abnormal and accidental states). To achieve this purpose an accident analysis of different design base accident (DBA) as loss of flow accident (LOFA), is required assessing reactor safety. The present work concerns this transients applied to Egypt Test and Research Reactor ETRR-3 (new reactor). An accident analysis code FLOWTR is developed to investigate the thermal behaviour of the core during such flow transients. The active core is simulated by two channels: 1 - hot channel (HC), and 2 - average channel (AC) representing the remainder of the core. Each channel is divided into four axial sections. The external loop, core plenums, and core chimney are simulated by different dynamic loops. The code includes modules for pump cast down, flow regimes, decay heat, temperature distributions, and feedback coefficients. FLOWTR is verified against results from RETRAN code, THERMIC code and commissioning tests for null transient case. The comparison shows a good agreement. The study indicates that for LOFA transients, provided the scram system is available, the core is shutdown safely by low flow signal (496.6 kg/s) at 1.4 s were the HC temperature reaches the maximum value, 45.64 o C after shutdown. On the other hand, if the scram system is unavailable, and at t = 47.33 s, the core flow decreases to 67.41 kg/s, the HC temperature increases to 164.02 o C, and the HC clad surface heat flux exceeds its critical value of 400.00 W/cm 2 resulting of fuel burnout. (author)

  6. Project requirements for reconstruction of the RA reactor ventilation system, Task 2.8. Measurement of radioactive iodine and other isotopes contents in the gas system of the RA reactor, Annex of the task

    International Nuclear Information System (INIS)

    Vujisic, Lj. et al

    1981-01-01

    This report is a supplement to the task 2.8. When planning and constructing the ventilation system, it was found that it is necessary to perform additional experiments during RA reactor operation at 2 MW power level for a longer period. In addition to the helium system, the potential source of radioactive pollutants is the space below the upper water shielding of the reactor. All the experimental and fuel channels are ending in this space. During repair and fuel exchange radioactivity can be released in this space. For that reason this space is important when planing and designing the filtration system for incidental conditions or planned dehermetisation of the reactor. The third point where radioactive isotope identification was done, was the entrance into the chimney during steady state operation and planned dehermetisation of the reactor. The following samples were measured: gas system during reactor operation at 2 MW power; entrance into the chimney during last 48 hours of reactor operation at 2 MW power; sample on the platform under the upper water shield with the opened fuel channel after the reactor shutdown; and simultaneously with the latter, measurement at the entrance to the chimney. This annex contains the list of identified radioactive isotopes, volatile and gaseous as well as concentration of volatile 131 I on the adsorbents [sr

  7. Reactor water level control device

    International Nuclear Information System (INIS)

    Utagawa, Kazuyuki.

    1993-01-01

    A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

  8. Decontamination and decommissioning project of the TRIGA Mark-2 and 3 research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jung, K J; Baik, S T; Chung, U S; Jung, K H; Park, S K; Lee, B J; Kim, J K; Yang, S H

    2000-01-01

    During the review on the decommissioning plan and environmental impact assessment report by the KINS, the number of the inquired items were two hundred and fifty one, and the answers were made and sent until September 10, 1999, as the screened review results were reported to Ministry of Science and Technology(MOST) in December 14, 1999, all the reviews on the licence were over. Radioactive liquid wastes of 400 tons generated during the operation of the research reactors including reactor vessels are stored in the facility of the research reactor 1 and 2. Those liquid wastes have the low-level-radioactivity which can be discharged to the surroundings, but was wholly treated to be vaporized naturally by means of the increased numbers of the natural vaporization disposal facilities with the annual capacity of 200 tons for the purpose of the minimized environmental contamination.

  9. Study on effects of development of reactor constant in fast reactor analysis

    International Nuclear Information System (INIS)

    Chiba, Gou

    2002-12-01

    Evaluation was carried out about an effect of development of the new generation reactor constant system that substitutes for the JFS library in fast reactor analysis. Analyzed cores were ZPPR in JUPITER critical experiment and several power reactor cores that were designed in the feasibility study. In the JUPITER analysis, large effects, over 10%, were observed in sodium void reactivity and sample Doppler reactivity. The former resulted from several factors, while the latter was due to an accurate of a resonance interaction effect between Doppler sample and core fuel. In the previous study, the effect had been evaluated in power reactor cores. The effect included an effect of corrosion of weighting spectrum because JFS-3-J3.2, which had been made with the incorrect weighting spectrum, was used in the evaluation. In the present study, JFS-3-J3.2R, which had been made with the correct weighting spectrum, was used. It was confirmed that the effect of development of reactor constant in power reactor was not as large as that in critical assembly. (author)

  10. Thermal and stress analyses of the reactor pressure vessel lower head of the Three Mile Island Unit 2

    International Nuclear Information System (INIS)

    Hashimoto, K.; Onizawa, K.; Kurihara, R.; Kawasaki, S.; Soda, K.

    1992-01-01

    Thermal and stress analyses were performed using the finite element analysis code ABAQUS to clarify the factors which caused tears in the stainless steel liner of the reactor pressure vessel lower head of the Three Mile Island Unit 2 (TMI-2) reactor pressure vessel during the accident on 28 March 1979. The present analyses covered the events which occurred after approximately 20 tons of molten core material were relocated to the lower head of the reactor pressure vessel. They showed that the tensile stress was highest in the case where the relocated core material consisting of homogeneous UO 2 debris was assumed to attack the lower head and the debris was then quenched. The peak tensile stress was in the vicinity of the welded zone of the penetration nozzle. This result agrees with the findings from the examination of the TMI-2 reactor pressure vessel that major tears in the stainless steel liner were observed around two penetration nozzles of the lower head. (author)

  11. The FR 2 reactor at Karlsruhe, F.R. Germany and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the FR 2 reactor and associated hot cell facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  12. Reactor physics measurements with 19-element ThOsub(2)-sup(235)UOsub(2) cluster fuel in heavy water moderator

    International Nuclear Information System (INIS)

    French, P.M.

    1985-02-01

    Low power lattice physics measurements have been performed with a single rod of 19-element thorium oxide fuel enriched with 1.45 wt. percent sub(235)UOsub(2) (93 percent enriched) in a simulated heavy water moderated and cooled power reactor core. The experiments were designed to provide data relevant to a power reactor irradiation and to obtain some basic information on the physics of uranium-thorium fuel material. Some theoretical flux calculations are summarized and show reasonable agreement with experiment

  13. The performance of ENDF/B-V.2 nuclear data for fast reactor calculations

    International Nuclear Information System (INIS)

    Atkinson, C.A.; Collins, P.J.

    1987-01-01

    Calculations with ENDF/B-V.2 data have been made for twenty-five fast-spectrum integral assemblies covering a wide range of sizes and compositions. Analysis was done by transport codes with refined cross section processing methods and detailed reactor modelling. The predictions of fission rate distributions and control rod worths were emphasized for the more prototypic benchmark cores. The results show considerable improvements in agreement with experiment compared with analysis using ENDF/B-IV data, but it is apparent that significant errors remain for fast reactor design calculations

  14. TMI-2 reactor vessel and balance of plant status

    International Nuclear Information System (INIS)

    Kuehn, G.A.

    1990-01-01

    In the fall of 1988 a corporate decision was made which concentrated effort on support of reactor vessel defueling and minimized activity in balance-of-plant areas. The auxiliary and fuel handling building are in a safe/stable state but final preparations for monitored storage won't be pursued until defueling and fuel shipping are complete. In addition to dispositioning fuel, the project is actively preparing for disposal of the Accident Generated Water (2.3 million gallons) by evaporation

  15. Abatement of fluorinated compounds using a 2.45 GHz microwave plasma torch with a reverse vortex plasma reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J.H.; Cho, C.H.; Shin, D.H. [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Oxikdo-dong, Gunsan-city, Jeollabuk-do (Korea, Republic of); Hong, Y.C., E-mail: ychong@nfri.re.kr [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Oxikdo-dong, Gunsan-city, Jeollabuk-do (Korea, Republic of); Shin, Y.W. [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Oxikdo-dong, Gunsan-city, Jeollabuk-do (Korea, Republic of); School of Advanced Green Energy and Environments, Handong Global University, Heunghae-eup, Buk-gu, Pohang-city, Gyeongbuk (Korea, Republic of)

    2015-08-30

    Highlights: • We developed a microwave plasma torch with reverse vortex reactor (RVR). • We calculated a volume fraction and temperature distribution of discharge gas and waste. • The performance of reverse vortex reactor increased from 29% to 43% than conventional vortex reactor. - Abstract: Abatement of fluorinated compounds (FCs) used in semiconductor and display industries has received an attention due to the increasingly stricter regulation on their emission. We have developed a 2.45 GHz microwave plasma torch with reverse vortex reactor (RVR). In order to design a reverse vortex plasma reactor, we calculated a volume fraction and temperature distribution of discharge gas and waste gas in RVR by ANSYS CFX of computational fluid dynamics (CFD) simulation code. Abatement experiments have been performed with respect to SF{sub 6}, NF{sub 3} by varying plasma power and N{sub 2} flow rates, and FCs concentration. Detailed experiments were conducted on the abatement of NF{sub 3} and SF{sub 6} in terms of destruction and removal efficiency (DRE) using Fourier transform infrared (FTIR). The DRE of 99.9% for NF{sub 3} was achieved without an additive gas at the N{sub 2} flow rate of 150 liter per minute (L/min) by applying a microwave power of 6 kW with RVR. Also, a DRE of SF{sub 6} was 99.99% at the N{sub 2} flow rate of 60 L/min using an applied microwave power of 6 kW. The performance of reverse vortex reactor increased about 43% of NF{sub 3} and 29% of SF{sub 6} abatements results definition by decomposition energy per liter more than conventional vortex reactor.

  16. Advances in reactor physics education: Visualization of reactor parameters

    International Nuclear Information System (INIS)

    Snoj, L.; Kromar, M.; Zerovnik, G.

    2012-01-01

    Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

  17. Slit-burst testing of cold-worked Zr-2.5 wt.% Nb pressure tubing for CANDU-PHW reactors

    International Nuclear Information System (INIS)

    Wilkins, B.J.S.; Barrie, J.N.; Zink, R.J.

    1978-12-01

    This report documents the available data on critical crack length of cold-worked Zr-2.5 wt.% Nb pressure tubing in CANDU reactors. In particular, it includes data for tubing removed from the Pickering 3 and 4 reactors. (author)

  18. Aspectos teóricos das políticas de cidadania: uma aproximação ao Bolsa Família

    Directory of Open Access Journals (Sweden)

    Walquiria Leão Rego

    2008-01-01

    Full Text Available O artigo visa a três objetivos. O primeiro, destacar muito rapidamente a história recente das teorias dos direitos e da cidadania. O segundo é uma tentativa de compreender o sentido das tensões existentes nas relações entre o Estado e a sociedade civil, a participação política, o mercado e a justiça distributiva. Este conjunto problemático será examinado no âmbito mais amplo da questão democrática. Finalmente, estabelecer as conexões entre as bases normativas de algumas teorias da cidadania para construir o quadro conceitual de uma pesquisa em curso sobre a atual política de transferência de renda focada nas pessoas mais pobres, pelo Programa Bolsa Família cujo foco são as mulheres pobres.This article has three objectives. The first is to point out very briefly the recent history of the rights and citizenship theories. The second is an attempt to understand the meaning of the tensions between the State and civil society, and political participation, market and distributive justice. This problematic whole is seen from the wider range of the democratic question. Finally, the article sets out the connections between the normative basis of some citizenship theories in order to build up a framework for a research in course about the Brazilian government income program (the "Family Grant" whose focus is poor women.

  19. Cronos 2: a neutronic simulation software for reactor core calculations; Cronos 2: un logiciel de simulation neutronique des coeurs de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Lautard, J J; Magnaud, C; Moreau, F; Baudron, A M [CEA Saclay, Dept. de Mecanique et de Technologie (DMT/SERMA), 91 - Gif-sur-Yvette (France)

    1999-07-01

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  20. TARMS, an on-line boiling water reactor operation management system. [3 D core simulator LOGOS 2

    Energy Technology Data Exchange (ETDEWEB)

    Iwamoto, T.; Sakurai, S.; Uematsu, H.; Tsuiki, M.; Makino, K.

    1984-12-01

    The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site tool for boiling water reactor core operation management. It was designed to support a complete function set to meet the requirement to the current on-line process computers. The functions can be divided into two categories. One is monitoring of the present core power distribution as well as related limiting parameters. The other is aiding site engineers or reactor operators in making the future reactor operating plan. TARMS performs these functions with a three-dimensional BWR core physics simulator LOGOS 2, which is based on modified one-group, coarse-mesh nodal diffusion theory. A method was developed to obtain highly accurate nodal powers by coupling LOGOS 2 calculations with the readings of an in-core neutron flux monitor. A sort of automated machine-learning method also was developed to minimize the errors caused by insufficiency of the physics model adopted in LOGOS 2. In addition to these fundamental calculational methods, a number of core operation planning aid packages were developed and installed in TARMS, which were designed to make the operator's inputs simple and easy.

  1. Reactor Core Internals Replacement of Ikata Units 1 and 2

    International Nuclear Information System (INIS)

    Ikeda, K.; Ishikawa, T.; Miyoshi, T.; Takagi, T.

    2012-01-01

    This paper presents an overview of the reactor core internals replacement project carried out at the Ikata Nuclear Power Station in Japan, which was the first of its kind among PWRs in the world. Failure of baffle former bolts was first reported in 1989 at Bugey 2 in France. Since then, similar incidents have been reported in Belgium and in the U.S., but not in Japan. However, the possibility of these bolts failing in Japanese plants cannot be denied in the future as operating hours increase. Ageing degradation mechanisms for the reactor core internals include irradiation-assisted stress corrosion cracking of baffle former bolts and mechanical wear of control rod guide cards. Two different approaches can be taken to address these ageing issues: to inspect and repair whenever a problem is found; and to replace the entire core internals with those of a new design having advanced features to prevent ageing degradation problems. The choice of our company was the latter. This paper explains the reasons for the choice and summarizes the replacement project activities at Ikata Units 1 and 2 as well as the improvements incorporated in the new design. (author)

  2. Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Khang, Ngo Phu [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    The Dalat nuclear research reactor was reconstructed from TRIGA MARK II reactor, built in 1963 with nominal power of 250 KW, and reached its planned nominal power of 500 kW for the first time in Feb. 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc. (author). 10 refs., 10 figs., 2 tabs.

  3. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

    Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

  4. Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

    International Nuclear Information System (INIS)

    D. Kokkinos

    2005-01-01

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

  5. Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Kalcheva, S [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Sikik, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-09-01

    BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water. The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).

  6. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  7. Applications: fission, nuclear reactors. Fission: the various ways for reactors and cycles

    International Nuclear Information System (INIS)

    Bacher, P.

    1997-01-01

    A historical review is presented concerning the various nuclear reactor systems developed in France by the CEA: the UNGG (graphite-gas) system with higher CO 2 pressures, bigger fuel assemblies and powers higher than 500 MW e, allowed by studies on reactor physics, cladding material developments and reactor optimization; the fast neutron reactor system, following the graphite-gas development, led to the Superphenix reactor and important progress in simulation based on experiment and return of experience; and the PWR system, based on the american license, which has been successfully accommodated to the french industry and generates up to 75% of the electric power in France

  8. A method of reactor power decrease by 2DOF control system during BWR power oscillation

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Suzuki, Katsuo

    1998-09-01

    Occurrence of power oscillation events caused by void feedback effects in BWRs operated at low-flow and high-power condition has been reported. After thoroughly examining these events, BWRs have been equipped with the SRI (Selected Rod Insertion) system to avoid the power oscillation by decreasing the power under such reactor condition. This report presents a power control method for decreasing the reactor power stably by a two degree of freedom (2DOF) control. Performing a numerical simulation by utilizing a simple reactor dynamics model, it is found that the control system designed attains a satisfactory control performance of power decrease from a viewpoint of setting time and oscillation. (author)

  9. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  10. Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

    Energy Technology Data Exchange (ETDEWEB)

    Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

    2016-06-15

    Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

  11. Catalytic combustion of the retentate gas from a CO2/H2 separation membrane reactor for further CO2 enrichment and energy recovery

    International Nuclear Information System (INIS)

    Hwang, Kyung-Ran; Park, Jin-Woo; Lee, Sung-Wook; Hong, Sungkook; Lee, Chun-Boo; Oh, Duck-Kyu; Jin, Min-Ho; Lee, Dong-Wook; Park, Jong-Soo

    2015-01-01

    The CCR (catalytic combustion reaction) of the retentate gas, consisting of 90% CO 2 and 10% H 2 obtained from a CO 2 /H 2 separation membrane reactor, was investigated using a porous Ni metal catalyst in order to recover energy and further enrich CO 2 . A disc-shaped porous Ni metal catalyst, namely Al[0.1]/Ni, was prepared by a simple method and a compact MCR (micro-channel reactor) equipped with a catalyst plate was designed for the CCR. CO 2 and H 2 concentrations of 98.68% and 0.46%, respectively, were achieved at an operating temperature of 400 °C, GHSV (gas-hourly space velocity) of 50,000 h −1 and a H 2 /O 2 ratio (R/O) of 2 in the unit module. In the case of the MCR, a sheet of the Ni metal catalyst was easily installed along with the other metal plates and the concentration of CO 2 in the retentate gas increased up to 96.7%. The differences in temperatures measured before and after the CCR were 31 °C at the product outlet and 19 °C at the N 2 outlet in the MCR. The disc-shaped porous metal catalyst and MCR configuration used in this study exhibit potential advantages, such as high thermal transfer resulting in improved energy recovery rate, simple catalyst preparation, and easy installation of the catalyst in the MCR. - Highlights: • The catalytic combustion of a retentate gas obtained from the H 2 /CO 2 separation membrane. • A disc-shaped porous nickel metal catalyst and a micro-channel reactor for catalytic hydrogen combustion. • CO 2 enrichment up to 98.68% at 400 °C, 50,000 h −1 and H 2 /O 2 ratio of 2.

  12. Reactor feedwater system

    International Nuclear Information System (INIS)

    Kagaya, Hiroyuki; Tominaga, Kenji.

    1993-01-01

    In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

  13. The impact of the Bolsa Família Program on food consumption: a comparative study of the southeast and northeast regions of Brazil.

    Science.gov (United States)

    Sperandio, Naiara; Rodrigues, Cristiana Tristão; Franceschini, Sylvia do Carmo Castro; Priore, Silvia Eloiza

    2017-06-01

    The aim of this study was to evaluate the impact of the Bolsa Família Program (PBF) on food consumption in the northeast and southeast regions of Brazil. The database was obtained from the individual food consumption module of the Household Budget Survey conducted in 2008-09. Consumption was assessed through two food records. The food was categorized into four groups: fresh or minimally processed food; culinary ingredients; processed food; and ultra-processed food. To analyze the impact, the propensity score matching method was used, which compares the individual recipients and non-recipients of the PBF in relation to a group of socioeconomic characteristics. After the propensity score was calculated, the impact of the PBF was estimated through the nearest-neighbor matching algorithm. In both regions, more than 60% of the daily total calories consumed by PBF recipients came from foods that had not undergone industrial processing. The recipients of PBF had a low level of consumption of processed and ultra-processed food in both regions, and an increased level of consumption of fresh or minimally processed food in the northeast. The results indicate the importance of adopting intersectoral policies in parallel to the PBF in order to strengthen healthy eating practices.

  14. Assessment of United States industry structural codes and standards for application to advanced nuclear power reactors: Appendices. Volume 2

    International Nuclear Information System (INIS)

    Adams, T.M.; Stevenson, J.D.

    1995-10-01

    Throughout its history, the USNRC has remained committed to the use of industry consensus standards for the design, construction, and licensing of commercial nuclear power facilities. The existing industry standards are based on the current class of light water reactors and as such may not adequately address design and construction features of the next generation of Advanced Light Water Reactors and other types of Advanced Reactors. As part of their on-going commitment to industry standards, the USNRC commissioned this study to evaluate US industry structural standards for application to Advanced Light Water Reactors and Advanced Reactors. The initial review effort included (1) the review and study of the relevant reactor design basis documentation for eight Advanced Light Water Reactors and Advanced Reactor Designs, (2) the review of the USNRCs design requirements for advanced reactors, (3) the review of the latest revisions of the relevant industry consensus structural standards, and (4) the identification of the need for changes to these standards. The results of these studies were used to develop recommended changes to industry consensus structural standards which will be used in the construction of Advanced Light Water Reactors and Advanced Reactors. Over seventy sets of proposed standard changes were recommended and the need for the development of four new structural standards was identified. In addition to the recommended standard changes, several other sets of information and data were extracted for use by USNRC in other on-going programs. This information included (1) detailed observations on the response of structures and distribution system supports to the recent Northridge, California (1994) and Kobe, Japan (1995) earthquakes, (2) comparison of versions of certain standards cited in the standard review plan to the most current versions, and (3) comparison of the seismic and wind design basis for all the subject reactor designs

  15. O Programa Institucional de Bolsas de Iniciação à Docência na articulação entre a experiência na escola e o ensino da universidade

    Directory of Open Access Journals (Sweden)

    Alcione Castro

    2016-06-01

    Full Text Available O artigo faz uma análise das contribuições do Programa Institucional de Iniciação à Docência através do Subprojeto de Aprendizagem e Assimilação Cooperativa vinculado ao Curso de Pedagogia Campus de Sinop – Mato Grosso. A pesquisa de caráter qualitativa foi desenvolvida na Escola Municipal de Educação Básica Lizamara Aparecida de Oliva Almeida, a coleta de dados ocorreu por meio de documentos e entrevistas semi-estruturadas com os bolsistas. Objetivou-se compreender o programa e as experiências que são mediadas pelos sujeitos como todo, tendo como embasamento teórico Paulo Freire. Nesta perspectiva, os resultados da experiência possibilitam qualificação do bolsista em formação. Palavras-chave: educação; programa de bolsas de iniciação à docência; Paulo Freire; pesquisa qualitativa.

  16. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    Jaouen, C.; Beroux, P.

    2012-01-01

    From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

  17. Model study of an automatic controller of the IBR-2 pulsed reactor

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Popov, A.K.

    2007-01-01

    For calculation of power transients in the IBR-2 reactor a special mathematical model of dynamics taking into account the discontinuous jump of reactivity by an automatic controller with the step motor is created. In the model the nonlinear dependence of the energy of power pulse on the reactivity and the influence of warming up of the reactor on the reactivity by means of introduction of a nonlinear feedback 'power-pulse energy - reactivity' are taken into account. With the help of the model the transients of relative deviation of power-pulse energy are calculated at various (random, mixed and regular) reactivity disturbances at the reactor mean power 1.475 MW. It is shown that to improve the quality of processes the choice of such regular values of parameters of the automatic controller is expedient, at which the least effect of smoothing of a signal acting on an automatic controller and the least speed of an automatic controller are provided, and the reduction of efficiency of one step of the automatic controller and introduction of a five-percent dead space are also expedient

  18. Backfitting of the FRG reactors

    Energy Technology Data Exchange (ETDEWEB)

    Krull, W [GKSS-Forschungszentrum Geesthacht GmbH, Geesthacht (Germany)

    1990-05-01

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U{sub 3}Si{sub 2} fuel. Both cooling towers were repaired. Replacement of instrumentation is planned.

  19. Backfitting of the FRG reactors

    International Nuclear Information System (INIS)

    Krull, W.

    1990-01-01

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U 3 Si 2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

  20. Compact stellarators as reactors

    International Nuclear Information System (INIS)

    Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

    2001-01-01

    Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

  1. MICROORGANISMOS ASOCIADOS A FRUTOS EMBOLSADOS DE GUAYABA TAIWANESA VARIEDAD TAI KUO

    Directory of Open Access Journals (Sweden)

    Rossy Morera-Montoya

    2009-01-01

    Durante la época lluviosa también se encontraron diferencias altamente significativas (X2= 0,04773 en la incidencia de microorganismos, teniendo como resultado: cuando control > bolsa de tergal > bolsa Taiwanesa. Los resultados sugieren que los microorganismos aislados de frutos enfermos podrían ser agentes responsables de la pérdida de frutas comercia les en las plantaciones tropicales de guayaba.

  2. Análise da relação risco e retorno em carteiras compostas por índices de bolsa de valores de países desenvolvidos e de países emergentes integrantes do bloco econômico BRIC Analysis of the relationship between risk and return in portfolios comprising stock exchange indexes from developed and emerging countries members of BRICs economic bloc

    Directory of Open Access Journals (Sweden)

    José Odálio dos Santos

    2010-12-01

    Full Text Available O objetivo deste trabalho foi analisar se a formação de carteiras de investimentos compostas por ativos internacionais pode proporcionar relações de risco e retorno mais vantajosas para o investidor. Paralelamente, analisou-se o estágio de integração entre as economias dos países selecionados por meio do modelo desenvolvido por Securato (1997, denominado Nível de Globalização Restritra (NGR. A pesquisa foi aplicada em dois períodos: 1996 a 2000, quando se intensifica a abertura de importantes mercados emergentes, e de 2003 a 2007 para a comparação dos resultados. Para analisar a contribuição da diversificação internacional, calculou-se o risco e o retorno de quatro carteiras assim formadas: 1. índices de bolsa de valores dos países desenvolvidos (Reino Unido, EUA e Japão e dos países que integram o BRIC; 2. índices de bolsa de valores dos EUA e dos países que integram o BRIC; 3. índices de bolsa de valores dos países que integram o BRIC e 4. índices de bolsa de valores dos países desenvolvidos. Os resultados empíricos sugerem que o investidor obteria melhores resultados, caso optasse por carteiras compostas pelos índices do mercado acionário dos Estados Unidos e dos países integrantes do BRIC. A adição desses ativos na carteira geraria menores índices de covariância, ou seja, a menor exposição de risco por unidade de retorno. Por outro lado, embora tenha aumentado o nível de globalização entre os mercados no período mais recente da pesquisa (2003-2007, constatou-se a necessidade de maior integração entre as economias dos países selecionados (NGR The aim of this study was to examine whether the formation of portfolios composed of international assets can provide risk and return more advantageous for the investor. In parallel, we analyzed the stage of integration between the economies of selected countries through the model developed by Securato (1997, called Restricted Globalization Level (NGR. The

  3. Design options for a bunsen reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Robert Charles

    2013-10-01

    This work is being performed for Matt Channon Consulting as part of the Sandia National Laboratories New Mexico Small Business Assistance Program (NMSBA). Matt Channon Consulting has requested Sandia's assistance in the design of a chemical Bunsen reactor for the reaction of SO2, I2 and H2O to produce H2SO4 and HI with a SO2 feed rate to the reactor of 50 kg/hour. Based on this value, an assumed reactor efficiency of 33%, and kinetic data from the literature, a plug flow reactor approximately 1%E2%80%9D diameter and and 12 inches long would be needed to meet the specification of the project. Because the Bunsen reaction is exothermic, heat in the amount of approximately 128,000 kJ/hr would need to be removed using a cooling jacket placed around the tubular reactor. The available literature information on Bunsen reactor design and operation, certain support equipment needed for process operation and a design that meet the specification of Matt Channon Consulting are presented.

  4. Relative neutronic performance of proposed high-density dispersion fuels in water-moderated and D2O-reflected research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Matos, J.E.; Snelgrove, J.L.

    1996-01-01

    This paper provides an overview of the neutronic performance of an idealized research reactor using several high density LEU fuels that are being developed by the RERTR program. High-density LEU dispersion fuels are needed for new and existing high-performance research reactors and to extend the lifetime of fuel elements in other research reactors. This paper discusses the anticipated neutronic behavior of proposed advanced fuels containing dispersions of U 3 Si 2 , UN, U 2 Mo and several uranium alloys with Mo, or Zr and Nb. These advanced fuels are ranked based on the results of equilibrium depletion calculations for a simplified reactor model having a small H 2 O-cooled core and a D 2 O reflector. Plans have been developed to fabricate and irradiate several uranium alloy dispersion fuels in order to test their stability and compatibility with the matrix material and to establish practical loading limits

  5. Relative neutronic performance of proposed high-density dispersion fuels in water-moderated and D2O-reflected research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Matos, J.E.; Snelgrove, J.L.

    1996-01-01

    This paper provides an overview of the neutronic performance of an idealized research reactor using several high density Leu fuels that are being developed by the Rarita program. High-density Leu dispersion fuels are needed for new and existing high-performance research reactors and to extend the lifetime of fuel elements in other research reactors. This paper discusses the anticipated neutronic behavior of proposed advanced fuels containing dispersions of U 3 Si 2 , UN, U 2 Mo and several uranium alloys with Mo, or Zr and Nb. These advanced fuels are ranked based on the results of equilibrium depletion calculations for a simplified reactor model having a small H 2 O-cooled core and a D 2 O reflector. Plans have been developed to fabricate and irradiate several uranium alloy dispersion fuels in order to test their stability and compatibility with the matrix material and to establish practical loading limits. (author)

  6. Performance analysis of photocatalytic CO2 reduction in optical fiber monolith reactor with multiple inverse lights

    International Nuclear Information System (INIS)

    Yuan, Kai; Yang, Lijun; Du, Xiaoze; Yang, Yongping

    2014-01-01

    Highlights: • A new optical fiber monolith reactor model for CO 2 reduction was developed. • Methanol concentration versus fiber location and operation parameters was obtained. • Reaction efficiency increases by 31.1% due to the four fibers and inverse layout. • With increasing space of fiber and channel center, methanol concentration increases. • Methanol concentration increases as the vapor ratio and light intensity increase. - Abstract: Photocatalytic CO 2 reduction seems potential to mitigate greenhouse gas emissions and produce renewable energy. A new model of photocatalytic CO 2 reduction in optical fiber monolith reactor with multiple inverse lights was developed in this study to improve the conversion of CO 2 to CH 3 OH. The new light distribution equation was derived, by which the light distribution was modeled and analyzed. The variations of CH 3 OH concentration with the fiber location and operation parameters were obtained by means of numerical simulation. The results show that the outlet CH 3 OH concentration is 31.1% higher than the previous model, which is attributed to the four fibers and inverse layout. With the increase of the distance between the fiber and the monolith center, the average CH 3 OH concentration increases. The average CH 3 OH concentration also rises as the light input and water vapor percentage increase, but declines with increasing the inlet velocity. The maximum conversion rate and quantum efficiency in the model are 0.235 μmol g −1 h −1 and 0.0177% respectively, both higher than previous internally illuminated monolith reactor (0.16 μmol g −1 h −1 and 0.012%). The optical fiber monolith reactor layout with multiple inverse lights is recommended in the design of photocatalytic reactor of CO 2 reduction

  7. MASTER-2.0: Multi-purpose analyzer for static and transient effects of reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Byung Oh; Song, Jae Seung; Joo, Han Gyu [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    MASTER-2.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the two group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM(Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with AFEN/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. Master-2.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P model can be used selectively. In addition, MASTER-2.0 is designed to cover various PWRs including SMART as well as WH-and CE-type reactors, providing all data required in their design procedures. (author). 39 refs., 12 figs., 4 tabs.

  8. Fast reactors worldwide

    International Nuclear Information System (INIS)

    Hall, R.S.; Vignon, D.

    1985-01-01

    The paper concerns the evolution of fast reactors over the past 30 years, and their present status. Fast reactor development in different countries is described, and the present position, with emphasis on cost reduction and collaboration, is examined. The French development of the fast breeder type reactor is reviewed, and includes: the acquisition of technical skills, the search for competitive costs and the spx2 project, and more advanced designs. Future prospects are also discussed. (U.K.)

  9. Application of stable adaptive schemes to nuclear reactor systems, (2)

    International Nuclear Information System (INIS)

    Kukuda, Toshio

    1979-01-01

    The parameter identification and adaptive control schemes applied in a previous study to a nonlinear point reactor are extended to the case of a loosely-coupled-core reactor with internal feedbacks, constituting a nonlinear overall system. Both schemes are shown to be stable, with the system newly represented on the pattern of the Model Reference Adaptive System (MRAS) with use made of the Lyapunov's method. For either parameter identification or adaptive control of a loosely-coupled-core reactor, there exists no canonical form of multiple input-multiple output system which can be directly applied for deriving the MRAS with the matrix version of the Kalman-Yakubovich lemma as it was in the case of the point reactor. This difficulty is circumvented by the practical assumption that the neutron density can be directly measured on each core as reactivity change is applied as input into the coupled core as a whole. For parameter identification, the model parameters are adaptively adjusted to those of each core, while for the adaptive control, plant parameters of each core can be adaptively compensated, again through control inputs, to asymptotically reduce the output error between the model and the plant. The point reactor is shown to correspond to a special case. (author)

  10. Upgradation of Apsara reactor

    International Nuclear Information System (INIS)

    Mammen, S.; Mukherjee, P.; Bhatnagar, A.; Sasidharan, K.; Raina, V.K.

    2009-01-01

    Apsara is a 1 MW swimming pool type research reactor using high enriched uranium as fuel with light water as coolant and moderator. The reactor is in operation for more than five decades and has been extensively used for basic research, radioisotope production, neutron radiography, detector testing, shielding experiments etc. In view of its long service period, it is planned to carry out refurbishment of the reactor to extend its useful life. During refurbishment, it is also planned to upgrade the reactor to a 2 MW reactor to improve its utilization and to upgrade the structure, system and components in line with the current safety standards. This paper gives a brief account of the design features and safety aspects of the upgraded Apsara reactor. (author)

  11. Análisis de la información que publica la Bolsa Mexicana de Valores: el caso de las empresas listadas en el mercado mexicano de derivados

    Directory of Open Access Journals (Sweden)

    Juan Fernando Guerrero Herrera

    2016-02-01

    Full Text Available Este trabajo presenta un análisis de la información publicada por la Bolsa Mexicana de valores (bmv respecto de las principales empresas listadas, así como de la información publicada por las propias empresas en sus respectivas páginas Web. El estudio incluye a las diez empresas más importantes que cotizan en la bmv. Entre dichas empresas se encuentran las que están listadas en el Mercado Mexicano de derivados (Mexder, es decir, empresas cuyas acciones son activos subyacentes para contratos de instrumentos financieros derivados. La información proporcionada es importante y de gran utilidad para inversionistas, analistas financieros y público en general, ya que contribuye al proceso de fijación de precios de activos financieros que son subyacentes en el Mexder. el presente análisis incluye cuadros con los contratos listados en el Mexder, tanto de futuros como de opciones, así como sus características principales. Asimismo, incluye tablas que muestran los volúmenes que actualmente operan los contratos e instrumentos referidos.

  12. Análise da contribuição do Programa Bolsa Família para o enfrentamento da pobreza e a autonomia dos sujeitos beneficiários

    Directory of Open Access Journals (Sweden)

    Maurício Gregianin Testa

    2013-12-01

    Full Text Available Esta pesquisa reflete sobre a contribuição do Programa Bolsa Família (PBF para o enfrentamento da pobreza e uma maior autonomia dos sujeitos beneficiários. Foram coletados dados quantitativos com 103 famílias beneficiárias, complementados com entrevistas qualitativas com profissionais e famílias. Entre as formas de privação (educação, saúde, trabalho e renda etc., a educação mostrou-se o aspecto de privação que obteve os melhores resultados na percepção das famílias na busca da autonomia. A participação em atividades de apoio social tem efeito direto na percepção de melhoria da situação da família; entretanto, a participação das famílias mostrou-se incipiente. Essas atividades contribuem para o desenvolvimento da autonomia e podem ser compreendidas como o principal mecanismo do programa para as pessoas encontrarem as "portas de saída".

  13. Mixed core management: Use of 93% and 72% enriched uranium in the BR2 reactor

    International Nuclear Information System (INIS)

    Ponsard, B.

    2000-01-01

    The BR2 reactor, put into operation in 1963 and refurbished from July 1995 till April 1997, is a 100 MW high-flux Materials Testing Reactor, using 93% 235 U enriched uranium as standard fuel, light water as coolant and beryllium as moderator. The present operating regime consists of five irradiation cycles per year at an operating power between 50 and 70 MW; each cycle is characterized by 21 days operation. In the framework of a 'qualification programme', six 72% 235 U fuel elements fabricated with uranium recovered from the reprocessing of BR2 spent fuel at UKAEA-Dounreay have been successfully irradiated in the period 1994-1995 reaching a maximum mean burnup of 48% without the release of fission products. Since 1998, this type of fuel element is irradiated routinely together with standard 93% 235 U fuel elements in order to optimize the utilization of the available HEU inventory. The purpose of this paper is to present the strategy developed in order to optimize the mixed core management of the BR2 reactor. (author)

  14. Effect of fuel assembly when changing from AFA 2G to AFA 3G on seismic loads of reactor internal

    International Nuclear Information System (INIS)

    Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

    2013-01-01

    Nonlinear seismic model for reactor with fuel assemblies of AFA 2G and AFA 3G is established. Using ANSYS software, seismic nonlinear time -history analysis is completed and the effects on seismic loads of reactor system are obtained. The result shows that when the fuel assembly changing from AFA 2G to AFA 3G, it is necessary to reevaluate the fuel assembly itself, but not the reactor internal. (authors)

  15. Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas V, G.; Banfield, J. [GE-Hitachi Nuclear Energy Americas LLC, Global Nuclear Fuel, Americas LLC, 3901 Castle Hayne Road, Wilmingtonm, North Carolina (United States); Avila N, A., E-mail: Gabriel.Cuevas-Vivas@ge.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

    2016-09-15

    This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

  16. Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

    International Nuclear Information System (INIS)

    Cuevas V, G.; Banfield, J.; Avila N, A.

    2016-09-01

    This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

  17. Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

    International Nuclear Information System (INIS)

    González Mantecón, Javier

    2015-01-01

    The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

  18. CO2 Energy Reactor - Integrated Mineral Carbonation: Perspectives on Lab-Scale Investigation and Products Valorization

    OpenAIRE

    Rafael M Santos; Pol CM Knops; Keesjan L Rijnsburger; Yi Wai eChiang

    2016-01-01

    To overcome the challenges of mineral CO2 sequestration, Innovation Concepts B.V. is developing a unique proprietary gravity pressure vessel (GPV) reactor technology and has focussed on generating reaction products of high economic value. The GPV provides intense process conditions through hydrostatic pressurization and heat exchange integration that harvests exothermic reaction energy, thereby reducing energy demand of conventional reactor designs, in addition to offering other benefits. In ...

  19. What occurred in the reactors

    International Nuclear Information System (INIS)

    Kudo, Kazuhiko

    2013-01-01

    Described is what occurred in the reactors of Fukushima Daiichi Nuclear Power Plant at the Tohoku earthquake and tsunami (Mar. 11, 2011) from the aspect of engineering science. The tsunami attacked the Plant 1 hr after the quake. The Plant had reactors in buildings no.1-4 at 10 m height from the normal sea level which was flooded by 1.5-5.5 m high wave. All reactors in no.1-6 in the Plant were the boiling water type, and their core nuclear reactions were stopped within 3 sec due to the first quake by control rods inserted automatically. Reactors in no.1-5 lost their external AC power sources by the breakdown and subsequent submergence (no.1-4) of various equipments and in no.1, 2 and 4, the secondary DC power was then lost by the battery death. Although the isolation condenser started to cool the reactor in no.1 after DC cut, its valve was then kept closed to heat up the reactor, leading to the reaction of heated Zr in the fuel tube and water to yield H 2 which was accumulated in the building: the cause of hydrogen explosion on 12th. The reactor in no.2 had the reactor core isolation cooling system (RCIC) which operated normally for few hrs, then probably stopped to heat up the reactor, resulting in meltdown of the core but no explosion occurred because of the opened door of the blowout panel on the wall by the blast of no.1 explosion. The reactor in no.3 had RCIC and high pressure coolant injection system, but their works stopped to result in the core damage and H 2 accumulation leading to the explosion on 14th. The reactor in no.4 had not been operated because of its periodical annual examination, but was explored on 15th, of which cause was thought to be due to backward flow of H 2 from no.3. Finally, the author discusses about this accident from the industrial aspect of the design of safety level (defense in depth) on international views, and problems and tasks given. (T.T.)

  20. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

    In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

  1. Catalytic wet oxidation of phenol in a trickle bed reactor over a Pt/TiO2 catalyst.

    Science.gov (United States)

    Maugans, Clayton B; Akgerman, Aydin

    2003-01-01

    Catalytic wet oxidation of phenol was studied in a batch and a trickle bed reactor using 4.45% Pt/TiO2 catalyst in the temperature range 150-205 degrees C. Kinetic data were obtained from batch reactor studies and used to model the reaction kinetics for phenol disappearance and for total organic carbon disappearance. Trickle bed experiments were then performed to generate data from a heterogeneous flow reactor. Catalyst deactivation was observed in the trickle bed reactor, although the exact cause was not determined. Deactivation was observed to linearly increase with the cumulative amount of phenol that had passed over the catalyst bed. Trickle bed reactor modeling was performed using a three-phase heterogeneous model. Model parameters were determined from literature correlations, batch derived kinetic data, and trickle bed derived catalyst deactivation data. The model equations were solved using orthogonal collocations on finite elements. Trickle bed performance was successfully predicted using the batch derived kinetic model and the three-phase reactor model. Thus, using the kinetics determined from limited data in the batch mode, it is possible to predict continuous flow multiphase reactor performance.

  2. Efecto de las condiciones de almacenamiento sobre la germinación de la semilla de dos patrones de cítricos

    Directory of Open Access Journals (Sweden)

    Jorge Herrera Q.

    2009-01-01

    Full Text Available Semillas de los patrones de cítricos Citrus volkameriana y citrumelo var. Swingle (Citrus paradisi x Poncirus trifoliata fueron almacenadas a 5ºC y tres contenidos de humedad: 17%, 20% y 23% en C. Volkameriana, y 24%, 27% y 30% en Swingle. Se utilizó tres tipos de empaque para almacenar la semilla: bolsa de polietileno doble, bolsa de papel dentro de bolsa de polietileno y bolsa de polietileno con arena con un contenido de humedad del 16%. Los resultados mostraron que Swingle perdió más rápidamente su germinación. Las semillas de C. volkameriana y Swingle conservaron su capacidad germinativa por mayor tiempo cuando se almacenaron a un 23% y un 24% de humedad, respectivamente. El tipo de empaque afectó la calidad fisiológica de las semillas, ya que C. volkameriana conservó una mayor germinación en bolsa de polietileno con arena, mientras que Swingle obtuvo sus mayores valores en bolsa de papel dentro de bolsa de polietileno.

  3. Investigation of sensors and instrument components in boiling water reactors. Results from Oskarshamn 2, Barsebaeck 2 in Sweden and Kernkraftwerk Muehleberg in Switzerland

    International Nuclear Information System (INIS)

    Bergdahl, B.G.

    1998-05-01

    The reactor monitoring instruments are important for the operation and safety of the plants. Static properties of the instruments are controlled annually, but the dynamic properties are rarely, if ever, examined. This study is the result of a project initiated by the Swedish Nuclear Power Inspectorate. The examinations are based on signal analysis and simultaneous measurement of multiple signals. Results from Oskarshamn 2 (O2), Barsebaeck 2 (B2) and Kernkraftwerk Muehleberg (KKM) are discussed in this report. The presentation is focused on reactor pressure and reactor level signals. the analysis of O2 revealed that the dynamics for 3 out of 14 sensors was 'filtered', meaning that a rapid level displacement is registered with delay. Inspection showed that a 1 sec filter was installed instead of 1.2 sec. The study also showed that old pressure-sensors in use both at O2 and B2 could not cope with high frequencies, and that some level-sensors were disturbed by mechanical oscillations at Bw. At KKM, a 2 Hz resonance was observed with 12 pressure and level sensors. The oscillation was created by an old pressure sensor and influenced the other sensors through the common impulse network

  4. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    International Nuclear Information System (INIS)

    Guigon, B.; Vacelet, H.; Dornbusch, D.

    2000-01-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs from activation analysis to power reactor fuel qualification. In this paper the main characteristics of the Jules Horowitz Reactor are presented. Safety criteria are explained. Finally, merits and disadvantages of UMo compared to the standard U 3 Si 2 fuel are discussed. (author)

  5. TMI-2 [Three Mile Island Unit 2] reactor building dose reduction task force

    International Nuclear Information System (INIS)

    Daniels, R.S.

    1988-01-01

    In late October 1982, the director of Three Mile Island Unit 2 (TMI-2) created the dose reduction task force with the objective of identifying the principal radiological sources in the reactor building and recommending actions to minimize the dose to workers on labor-intensive projects. Members of the task force were drawn form various groups at TMI. Findings and recommendations were presented to the US Nuclear Regulatory Commission in a briefing on November 18, 1982. The task force developed a three-step approach toward dose reduction. Step 1 identified the radiological sources. Step 2 modeled the source and estimated its contribution to the general area dose rates. Step 3 recommended actions to achieve dose reductions consistent with general exposure rate goals

  6. A novel auto-thermal reforming membrane reactor for high purity H2

    International Nuclear Information System (INIS)

    Tony Boyd; Grace, J.R.; Lim, C.J.; Adris, A.M.

    2006-01-01

    A novel hydrogen reactor based on steam reforming of natural gas has been developed and tested. The reactor produces high purity hydrogen using in-situ perm-selective membranes installed in a fluidized catalyst bed, thus shifting the thermodynamic equilibrium of the SMR reaction and eliminating the need for downstream hydrogen purification. The reactor is particularly suited to auto-thermal reforming, where air is added to the reformer to provide the endothermic reaction heat, thus eliminating the need to indirectly heat the reactor. The gas flow pattern within the fluidized bed induces an internal circulation of catalyst particles between the central SMR reaction (permeation) zone and an outer annulus. The circulating hot catalyst particles from the oxidation zone carry the required endothermic heat of reaction for the reforming, while ensuring that the palladium membranes are not exposed to excessive temperatures or to oxygen. Another beneficial characteristic of the reactor is that very little of the nitrogen present in the oxidation air reaches the reaction zone, thus maintaining the hydrogen driving force for the perm-selective membranes. Pilot plant results carried out in a semi-industrial scale reactor will be presented. The reactor was operated up to 650 C and 14 bar. Pure hydrogen (99.999+%) was initially obtained from the reactor and an equilibrium shift was demonstrated. (authors)

  7. Operating US power reactors

    International Nuclear Information System (INIS)

    Silver, E.G.

    1988-01-01

    This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of September 30, 1987, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (July, August, and September 1987) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. In addition to the tabular data, this article discusses other significant occurrences and developments that affected licensed US power reactors during this reporting period. Status changes at Braidwood Unit 1, Nine Mile Point 2, and Beaver Valley 2 are discussed. Other occurrences discussed are: retraining of control-room operators at Peach Bottom; a request for 25% power for Shoreham, problems at Fermi 2 which delayed the request to go to 75% power; the results of a safety study of the N Reactor at Hanford; a proposed merger of Pacific Gas and Electric with Sacramento Municipal Utility District which would result in the decommissioning of Rancho Seco; the ordered shutdown of Oyster Creek; a minor radioactivity release caused by a steam generator tube rupture at North Anna 1; and 13 fines levied by the NRC on reactor licensees

  8. Necessity of research reactors

    International Nuclear Information System (INIS)

    Ito, Tetsuo

    2016-01-01

    Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

  9. Conceptual design of reactor assembly of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Selvaraj, A.; Balasubramaniyan, V.; Raghupathy, S.; Elango, D.; Sodhi, B.S.; Chetal, S.C.; Bhoje, S.B.

    1996-01-01

    The conceptual design of Reactor Assembly of 500 MWe Prototype Fast Breeder Reactor (as selected in 1985) was reviewed with the aim of 'simplification of design', 'Compactness of the reactor assembly' and 'ease in construction'. The reduction in size has been possible by incorporating concentric core arrangement, adoption of elastomer seals for Rotatable plugs, fuel handling with one transfer arm type mechanism, incorporation of mechanical sealing arrangement for IHX at the penetration in Inner vessel redan and reduction in number of components. The erection of the components has been made easier by adopting 'hanging' support for roof slab with associated changes in the safety vessel design. This paper presents the conceptual design of the reactor assembly components. (author). 8 figs, 2 tabs

  10. Environmental assessment for decontamination of the Three Mile Island Unit 2 reactor building atmosphere. Addendum 2. Draft NRC staff report for public comment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-04-01

    The reactor building purge system is an existing system originally installed for purging the reactor building atmosphere during normal operation or maintenance conditions. Use of the reactor building purge system in conjunction with the hydrogen control subsystem evaluated in Section 6.1 represents a variation in the purging alternative for decontaminating the Unit 2 reactor building atmosphere. This variation in the purging alternative would function only under meteorological conditions favorable to atmospheric dispersion. The reactor building purge system is capable of purging the building at flow rates of 5,000-50,000 cfm. Actual purge rates authorized during any time interval would be dependent on meteorological conditions and reactor building concentrations. Like the hydrogen control subsystem, this system would remove reactor building atmosphere through a filter system and discharge it through the 160-ft plant vent stack to the environment. The advantage of using the reactor building purge system in conjunction with the hydrogen control system is that it could decontaminate the reactor building atmosphere in a total elapsed purge time as short as approximately 5 days, as compared with the 60 days that would be required if the hydrogen purge subsystem were used alone. Use of this variation in the purge alternative would result in the release of radioactive materials to the environment. However, calculations based on actual meteorological and release-rate data would be used to monitor radioactive releases so that they do not exceed the requirements of 10 CFR Part 20, the design objectives of 10 CFR Part 50, Appendix 1 and the applicable requirements of 40 CFR 190.10.

  11. Re criticality assessment following reactor core damage in Fukushima unit 2

    International Nuclear Information System (INIS)

    Jeong, Hae Sun; Song, Jin Ho; Park, Chang Je; Ha, Kwang Soon; Song, Yong Mann; Ryu, Eun Hyun

    2012-01-01

    Following the severe core damage accident at the Fukushima nuclear power plants (NPPs), many researchers have studied a possible Re criticality caused by core melting or corium. However, no one can accurately examine the internal conditions of the reactor vessel, and thus there have been different opinions from some organizations depending on their assumption and analysis methods. If there is a potential Re criticality in the reactor vessel, some counter plans for the accident management should be established to prevent and mitigate re criticality, and to return the plant to a safe and stable state. In this study, the criticality level following a severe core damage accident was first analyzed using the MCNPX 2.6.0 code. Based on this result, practical strategies in terms of accident management were obtained by charging soluble boron (H 3B O 3) into re flooded water

  12. RB reactor as the U-D2O benchmark criticality system

    International Nuclear Information System (INIS)

    Pesic, M.

    1998-01-01

    From a rich and valuable database fro 580 different reactor cores formed up to now in the RB nuclear reactor, a selected and well recorded set is carefully chosen and preliminarily proposed as a new uranium-heavy water benchmark criticality system for validation od reactor design computer codes and data libraries. The first results of validation of the MCNP code and adjoining neutron cross section libraries are resented in this paper. (author)

  13. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 2. User's manual

    International Nuclear Information System (INIS)

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear energy reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 2: User's Manual) describes the input requirements of VIPRE and its auxiliary programs, SPECSET, ASP and DECCON, and lists the input instructions for each code

  14. Cadmium-emitter self-powered thermal neutron detector performance characterization & reactor power tracking capability experiments performed in ZED-2

    Energy Technology Data Exchange (ETDEWEB)

    LaFontaine, M.W., E-mail: physics@execulink.com [LaFontaine Consulting, Kitchener, Ontario (Canada); Zeller, M.B. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Nielsen, K. [Royal Military College of Canada, SLOWPOKE-2 Reactor, Kingston, Ontario (Canada)

    2014-07-01

    Cadmium-emitter self-powered thermal neutron flux detectors (SPDs), are typically used for flux monitoring and control applications in low temperature, test reactors such as the SLOWPOKE-2. A collaborative program between Atomic Energy of Canada, academia (Royal Military College of Canada (RMCC)) and industry (LaFontaine Consulting) was initiated to characterize the incore performance of a typical Cd-emitter SPD; and to obtain a definitive measure of the capability of the detector to track changes in reactor power in real time. Prior to starting the experiment proper, Chalk River Laboratories' ZED-2 was operated at low power (5 watts nominal) to verify the predicted moderator critical height. Test measurements were then performed with the vertical center of the SPD emitter positioned at the vertical mid-plane of the ZED-2 reactor core. Measurements were taken with the SPD located at lattice position L0 (near center), and repeated at lattice position P0 (in D{sub 2}O reflector). An ionization chamber (part of the ZED-2 control instrumentation) monitored reactor power at a position located on the south side of the outside wall of the reactor's calandria. These experiments facilitated measurement of the absolute thermal neutron sensitivity of the subject Cd-emitter SPD, and validated the power tracking capability of said SPD. Procedural details of the experiments, data, calculations and associated graphs, are presented and discussed. (author)

  15. Feedback from dismantling operations (level 2) on EDF's first generation reactors

    International Nuclear Information System (INIS)

    West, J P.; Dionisio-Gomes, A.; Kus, J P.; Mervaux, P.; Bernet, P.; Dalmas, R.

    2003-01-01

    EDF's policy as regards the dismantling of the reactors that have ceased commercial operation, namely the eight power plants of the first generation and the Creys-Malville power plant, is explained. Generally speaking, prior to the year 2001, EDF had opted for the de-construction of these power plants to comply with a 'long wait' scenario, which consisted of waiting for a period of 5 to 10 years to achieve IAEA level 2 (partial release of the site), then postponing the total de-construction of the facilities for 25 to 50 years. Today, EDF has decided to undertake the total de-construction of these reactors, which have ceased commercial operation, over a period of 25 years. The purpose of this document is to present: - The reactors concerned, their background and their 'regulatory' situation, - The main operations performed and/or currently in progress, - The main elements of feedback from such operations, shedding light on the approach adopted in 2001. The installations concerned by the de-construction programme are as follows: - The 8 power plants of the first generation, which were built during the fifties and sixties and ceased commercial operation between 1973 and 1994, namely: Brennilis (industrial prototype using heavy water technology, jointly operated by EDF and CEA), the 6 power units of the NUGG type (natural uranium gas graphite) at Chinon, Saint-Laurent des Eaux and Bugey and the PWR reactor at Chooz A, - The storage silos at Saint-Laurent, where the sleeves for the fuel assemblies of reactors SLA1 and SLA2 are stored, corresponding to approximately 2000 tonnes of graphite, - The Creys-Malville reactor, FBR (fast breeder reactor) shut down in accordance with a government decision, which is currently undergoing decommissioning. At the current stage, our feedback from the dismantling operations carried out on nuclear facilities is based on (i) the work carried out or in progress that will make it possible to achieve the equivalent of IAEA level 2 in the

  16. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'JOYO'. Recovery of MARICO-2 sample part

    International Nuclear Information System (INIS)

    Ashida, Takashi; Ito, Hideaki

    2015-01-01

    At Joyo reactor MK-III core in May 2007, due to the design deficiencies of the disconnect mechanism of the holding part and the sample part of the experimental apparatus with instrumentation lines (MARICO-2), a disconnect failure incident occurred in the sample part after irradiation test. The deformation of the sample part due to this failure incurred its interference with the lower surface of reactor core upper structure and the holddown axis body. By this, the operating range of the rotary plug was restricted, leading to the partial inhibition of the fuel exchange function that precluded the access to 1/4 of the assemblies of the reactor core. In face of restoration work, the preparation for restoration such the exchange of upper core structure, and the recovery of MARICO-2 sample part are under way. The following items are introduced here: (1) summary of restoration work and overall process of restoration work, (2) recovery operation of MARICO-2 sample part, (3) exchange of the upper core structure that was conducted this year, and (4) results of recovery of MARIKO-2 sample part. (A.O.)

  17. A thermal hydraulic analysis in PWR reactors with UO2 or (U-Th)O2 fuel rods employing a simplified code

    International Nuclear Information System (INIS)

    Santos, Thiago A. dos; Maiorino, José R.; Stefanni, Giovanni L. de

    2017-01-01

    In order to project a nuclear reactor, the neutronic calculus must be validated, so that its thermal limits and safety parameters are respected. Considering this issue, this research aims to evaluate the APTh-100 reactor thermal limits. This PWR is a project developed in Universidade Federal do ABC (UFABC) using fuel composed of Uranium and Thorium oxide mixed (U,Th)O 2 . For this purpose, a simplified, although conservative, code was developed in a MATLAB environment named STC-MOX-Th 'Simplified Thermal-hydraulics Code-Mixed Oxide Thorium'. This code provides axial and radial temperature distribution, as well as DNBR distribution over the hottest channel of the reactor core. Moreover, it brings other hydraulic quantities, such as pressure drop over the fuel rod, considering any fuel proportion of (U,Th)O 2 .The software uses basic laws of conservation of mass, momentum and energy, it also calculates the thermal conduction equation, considering the thermal conductive coefficient as a temperature function. In order to solve this equation, the finite elements method was used. Furthermore, the proportion of 36% of UO 2 was used to evaluate the temperature over the fuel rod and DNBR minimum in three burn conditions: beginning, middle and ending. The program has proven to be efficient in every condition and the results evidenced that the APTh-1000 reactor, in an initial analysis, has its thermal limits within the recommended security parameters. (author)

  18. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  19. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

    A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

  20. Criticality analysis for mixed thorium-uranium fuel in the Angra-2 PWR reactor using KENO-VI

    Energy Technology Data Exchange (ETDEWEB)

    Wichrowski, Caio C.; Gonçalves, Isadora C.; Oliveira, Claudio L.; Vellozo, Sergio O.; Baptista, Camila O., E-mail: wichrowski@ime.eb.br, E-mail: isadora.goncalves@ime.eb.br, E-mail: d7luiz@yahoo.com.br, E-mail: vellozo@ime.eb.br, E-mail: camila.oliv.baptista@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Seção de Engenharia Nuclear

    2017-07-01

    The increasing energy demand associated to the current sustainability challenges have given the thorium nuclear fuel cycle renewed interest in the scientific community. Studies have focused on energy production in different reactor designs through the fission of uranium 233, the product of thorium fertilization by neutrons. In order to make it possible for near future applications a strategy based on the adaptation of current nuclear reactors for the use of thorium fuels is being considered. In this work, bearing in mind these limitations, a code was used to evaluate the effect on criticality (k{sub inf}) of the mixing of thorium and uranium in different proportions in the fuel of a PWR, the German designed Angra-2 Brazilian reactor in order to scrutinise its behaviour and determine the feasibility of an adapted ThO{sub 2}-UO{sub 2} mixed fuel cycle using current PWR technology. The analysis is performed using the KENO-VI module in the SCALE 6.1 nuclear safety analysis simulation code and the information is taken from the Angra-2 FSAR (Final Security Analysis Report). (author)