WorldWideScience

Sample records for fuel type guia

  1. Guias ópticos para sensores

    OpenAIRE

    Marcos de Castro Pacitti

    1993-01-01

    Neste trabalho foram estudados guias ópticos para atuarem como sensores ou em sistemas sensores. Ênfase especial foi dada em guias ópticos retangulares, para dispositivos ópticos integrados, e fibras ópticas elípticas. A analise de fibras ópticas elípticas resultou em um modelamento pratico para projeto das mesmas em aplicações onde se deseja operar com dois modos guiados, situação esta que se mostra muito atraente para implementação de diversos tipos de sensores a fibra óptica. A seguir es...

  2. Econlit (EBSCOHost): guia d'ús. Gener 2011

    OpenAIRE

    Universitat de Barcelona. CRAI

    2011-01-01

    Guia de la base de dades Econlit sobre economia i finances produïda per l'American Economic Association (AEA) i distribuïda per EBSCOHost. És la font més important del món en referències a la literatura econòmica.

  3. Perfil vocal do guia de turismo Vocal profile of tourism guide

    Directory of Open Access Journals (Sweden)

    Elisângela Barros Soares

    2006-12-01

    Full Text Available OBJETIVO: caracterizar o perfil vocal dos guias de turismo, bem como gênero e idade. MÉTODOS: participaram desse estudo 23 guias de turismo, de ambos os gêneros, com idade entre 25 a 64 anos, participantes do Sindicato de Guias de Turismo do Estado de Pernambuco, que compareceram às reuniões trimestrais no período da coleta. Trata-se de um estudo de caráter descritivo, observacional e transversal. Para coleta foi realizada avaliação perceptivo-auditiva GRBAS. RESULTADOS: observou-se que a maioria dos guias apresentou loudness adequada, pitch normal e voz alterada. Além disso, as médias dos tempos máximos de fonação das vogais e das fricativas encontravam-se reduzidas e ataque vocal isocrônico. A ressonância, na maioria dos guias, estava equilibrada, mas houve uma incidência de ressonância laringo-faringea. A articulação foi precisa, com tipo e modo respiratório misto e nasal, respectivamente. Quanto à escala GRBAS as alterações apareceram de forma leve no G (grau de alteração vocal em 68%. CONCLUSÃO: na amostra estudada, a maioria era do gênero feminino com média de idade de 46 anos, e perfil vocal caracterizado por tempo máximo de fonação reduzidos, relação s/z adequado, ataque vocal isocrônico, pitch normal, loudness adequado, qualidade vocal alterada, com presença de rouquidão, soprosidade, tensão. A ressonância da maioria estava equilibrada e a articulação precisa, com tipo e modo respiratório misto e nasal, respectivamente. Quanto à escala GRBAS, as alterações apareceram de forma leve no grau de alteração vocal (G em 68% e tensão (S em 78% dos sujeitos.PURPOSE: to characterize the vocal profile of tourism guides, as well as gender and age. METHODS: 23 guides took part in this study, of both genders, with age between 25 to 64 years, partakers of the Union of Tourism Guides of the State of Pernambuco, who appeared to the quarterly meetings in the period of the collection. It is a descriptive

  4. Tema 8. Principis físics dels semiconductors (Guia del tema)

    OpenAIRE

    Beléndez Vázquez, Augusto

    2011-01-01

    Guia del "Tema 8. Principis físics dels semiconductors" de l'assignatura "Fonaments Físics de l'Enginyeria I" de "Grau en Enginyeria en So i Imatge" impartit a l'Escola Politècnica Superior de la Universitat d'Alacant.

  5. The Tourist Attractions in Curitiba - PR: An overview by Guia Brazil Quatro Rodas

    Directory of Open Access Journals (Sweden)

    Franciele Cristina Manosso

    2015-04-01

    Full Text Available The tour guides books can be worked as relevant instruments for the dissemination of destinations, because encompass information about the destinations' attractions, equipment, infrastructure and tourism services. Then, the tourist can through this, conceiving a destination’s image that will visit, because with the elements worked by the guides, they can get a comprehensive view of what is offered by the destination. Therefore, this article has as objective the analysis of the tourist attractions in Curitiba’s city, according to Guia Brazil Quatro Rodas, presented in the years 1989, 2001, 2013. Thus, the paper aims to also highlight how the document presents the city to its visitors, analyzing the description and the photographs that are showing Curitiba to the tourist. For this, was used as methodology the literature's and documents' research, the latter being applied to observe which attractions, description, and photos were used in the dissemination of Curitiba in the Guia Brazil. As the study results, it can be noted that there was an increase in the number of tourist attractions in Curitiba presented by Guia Brazil Quatro Rodas, and some of these have progressed in the respect of their classification, in the other words, they have improved in quality. Another issue is the publication’s evolution in question the destinations’ presentation to the reader, because more information came to be published, and also began to be used photos to show to the tourist the Curitiba destination.

  6. Adaptação multicêntrica do guia para a gestão autônoma da medicação

    Directory of Open Access Journals (Sweden)

    Rosana Teresa Onocko Campos

    2012-12-01

    Full Text Available O uso crescente de psicofármacos e o baixo empowerment dos usuários mostram-se críticos à qualificação da assistência em Saúde Mental no Brasil. Este estudo, abrangendo três cidades brasileiras, objetivou a elaboração do Guia Brasileiro da Gestão Autônoma da Medicação (GGAM-BR, com base na tradução e adaptação de guia desenvolvido no Canadá; e a avaliação dos efeitos do uso do GGAM-BR na formação de trabalhadores de saúde mental. Constituíram-se grupos de intervenção (GIs para compartilhamento das experiências com tratamento medicamentoso, a partir dos temas propostos no guia; e foram realizados grupos focais antes e após os GIs. Importantes mudanças em relação ao texto original do guia Canadense foram implementadas, levando em conta a realidade brasileira. Constatou-se que o GGAM-BR constitui estratégia potente de fomento à participação ativa dos usuários na gestão do tratamento e do serviço, incidindo positivamente na formação de trabalhadores.

  7. Avaliação das guias condilar e incisal em função da curva de compensação e da altura das cúspides - Releitura das Leis de Articulação de Hanau

    OpenAIRE

    Roger Nishyama

    2011-01-01

    Este estudo teve como objetivo avaliar a variação das guias sagitais (guia condilar e guia incisal), em função do plano de orientação (curvatura e inclinação) e da altura das cúspides, no movimento protrusivo mandibular, in vitro, buscando as relações entre esses fatores, tanto de forma analógica como digital. No estudo analógico foram avaliadas as inclinações das guias condilares, inclinações da guia incisal, proeminência da curva de compensação e inclinação do plano oclusal. Para isso foram...

  8. O que visitar em Paris durante a Exposição Universal de 1878: um guia turístico para geólogos

    Directory of Open Access Journals (Sweden)

    Maria Margaret Lopes

    2015-10-01

    Full Text Available Este artigo parte da tradição constituída dos guias de turismo, para comentar os guias elaborados para as Exposições Universais. Apresenta um exemplo específico desses guias de turismo científico preparado para um Congresso que se realizou no âmbito de uma Exposição e que se tornaria modelar para os próximos eventos da área. Trata-se do Guide du géologue à l’Exposition universelle de 1878 et dans les collections publiques et privées de Paris. O Guide foi organizado para orientar os geólogos estrangeiros que participaram do Primeiro Congresso Internacional de Geologia realizado durante a Exposição Universal de Paris de 1878.

  9. Desempenho de modelos de calibração de guias de onda acopladas a TDR e a multiplexadores em três tipos de solos Performance of calibration models for TDR and multiplexer - connected waveguides in three soil types

    Directory of Open Access Journals (Sweden)

    Eugenio Ferreira Coelho

    2006-02-01

    Full Text Available O trabalho teve como objetivo avaliar modelos de calibração para dois tipos de guias de onda de TDR, referentes a dois equipamentos (Trase System e TDR 100, acopladas diretamente ao analisador de umidade ou a multiplexadores. Amostras de três tipos de solo foram acondicionadas em segmentos de tubos de PVC e saturadas. Dois tipos de guias de onda de três hastes, com capacitor e com resistor foram inseridas dentro de cada segmento de tubo com solo e conectadas a dois equipamentos de TDR, diretamente no testador de cabos ou via multiplexadores. Dados de umidade obtidos por gravimetria e da constante dielétrica foram tomados em cada coluna durante a secagem do solo da saturação até umidades próximas do limite inferior de disponibilidade de água por meio de leituras com as guias de onda conectadas ao testador de cabos e conectadas ao multiplexador. Um modelo polinomial cúbico foi ajustado aos dados da constante dielétrica do solo (épsilon e da correspondente umidade (teta e cinco modelos de determinação de q em função de e foram testados quanto ao desempenho. Os resultados mostraram que não houve diferença significativa na calibração das guias de onda com capacitor para uso com a TDR Trase System, considerando a conexão das guias ao analisador de umidade ou a multiplexadores. No caso da TDR 100, as guias de onda com resistor devem ser calibradas conforme o seu uso. O modelo cúbico foi o de melhor desempenho seguido pelo modelo de Roth que estimou, com boa exatidão, os valores da constante dielétrica e da umidade com a mais próximo de 0,5 para as guias de onda com capacitor que com as guias com resistor.The study aimed at evaluating of calibration models for two kinds of TDR waveguides used with Trase System and TDR 100 equipments linked to the cable tester or the multiplexer. Disturbed samples of three soils were packed in PVC columns and, after soil saturation, two TDR waveguides of three rods with capacitor and with

  10. Food Science and Technology Abstracts (FSTA): guia ràpida. Gener 2011

    OpenAIRE

    Universitat de Barcelona. CRAI

    2011-01-01

    Guia ràpida de la base de dades bibliogràfica (FSTA) d'àmbit mundial sobre ciència, tecnologia i química alimentària, nutrició i salut humana, biotecnologia i toxicologia. Buida prop de 1800 revistes especialitzades, monografies, conferències, tesis, patents, legislació, etc. publicats en unes 40 llengües.

  11. Fuel loads and fuel type mapping

    Science.gov (United States)

    Chuvieco, Emilio; Riaño, David; Van Wagtendonk, Jan W.; Morsdof, Felix; Chuvieco, Emilio

    2003-01-01

    Correct description of fuel properties is critical to improve fire danger assessment and fire behaviour modeling, since they guide both fire ignition and fire propagation. This chapter deals with properties of fuel that can be considered static in short periods of time: biomass loads, plant geometry, compactness, etc. Mapping these properties require a detail knowledge of vegetation vertical and horizontal structure. Several systems to classify the great diversity of vegetation characteristics in few fuel types are described, as well as methods for mapping them with special emphasis on those based on remote sensing images.

  12. Guia de receitas brasileiras: uma saborosa viagem pela literatura

    OpenAIRE

    Santos, Denísia Moraes dos

    2011-01-01

    Esta dissertação tem por objetivo compreender, sob o ponto de vista da Análise Dialógica do Discurso, o processo de construção dos sentidos em Guia de receitas brasileiras ─ texto premiado na primeira edição do Concurso Cultural Viagem Nestlé pela Literatura (1999). Com o tema A literatura e a vida nos 500 anos de Brasil , essa edição desafiou alunos e professores a produzir um intertexto literário-cultural, utilizando cinco obras da literatura brasileira ─ Cronistas do Descobrim...

  13. New type fuel exchange system

    International Nuclear Information System (INIS)

    Meshii, Toshio; Maita, Yasushi; Hirota, Koichi; Kamishima, Yoshio.

    1988-01-01

    When the reduction of the construction cost of FBRs is considered from the standpoint of the machinery and equipment, to make the size small and to heighten the efficiency are the assigned mission. In order to make a reactor vessel small, it is indispensable to decrease the size of the equipment for fuel exchange installed on the upper part of a core. Mitsubishi Heavy Industries Ltd. carried out the research on the development of a new type fuel exchange system. As for the fuel exchange system for FBRs, it is necessary to change the mode of fuel exchange from that of LWRs, such as handling in the presence of chemically active sodium and inert argon atmosphere covering it and handling under heavy shielding against high radiation. The fuel exchange system for FBRs is composed of a fuel exchanger which inserts, pulls out and transfers fuel and rotary plugs. The mechanism adopted for the new type fuel exchange system that Mitsubishi is developing is explained. The feasibility of the mechanism on the upper part of a core was investigated by water flow test, vibration test and buckling test. The design of the mechanism on the upper part of the core of a demonstration FBR was examined, and the new type fuel exchange system was sufficiently applicable. (Kako, I.)

  14. Desempenho de diferentes guias de ondas para uso com o analisador de umidade TRASE Performance of different waveguides for use with the TRASE water content analyser

    Directory of Open Access Journals (Sweden)

    Eugênio F. Coelho

    2001-04-01

    Full Text Available O trabalho teve como objetivo avaliar a relação entre a umidade e a constante dielétrica aparente do solo e diferentes guias de onda para uso com o analisador de umidade Trase, que opera de acordo com o princípio da reflectometria no domínio do tempo - TDR. Amostras indeformadas e deformadas de duas manchas de textura diferente de um Latossolo Amarelo Distrófico foram retiradas do campo e acondicionadas em recipientes de 10 L, perfazendo quatro repetições para cada textura e estrutura. Foram construídas três diferentes guias de onda, com hastes de 0,15 m, sendo uma com capacitor e uma sem capacitor no início da guia, com espaçamento de 0,009 m entre hastes e outra sem capacitor, com espaçamento de 0,022 m entre hastes. Essas guias de onda, juntamente com guias originais do fabricante com hastes de 0,20 m, espaçadas 0,022 m, foram inseridas individualmente em cada recipiente. Dados de umidade obtidos gravimetricamente e pelo analisador TRASE, e da constante dielétrica, foram tomados usando-se todas as guias de onda em cada recipiente durante a secagem do solo, de 0,35 m³ m-3 a 0,10 m³ m-3. Três modelos matemáticos foram ajustados aos dados de umidade e da correspondente constante dielétrica do solo gerados pelas guias de onda do fabricante. Um modelo exponencial foi considerado como mais adequado para estimativas dos teores de água, em função da constante dielétrica para a guia de onda do fabricante. Todas as guias de onda avaliadas apresentam viabilidade de uso, desde que previamente calibradas.The objective of this work was to evaluate the relations between soil water content and the soil bulk dielectric constant, and to study different waveguides of a TRASE soil water content analyzer that operates according to TDR principles. Non-destructive and destructive samples of two sites of different texture of a Dystrophic Yellow Latossol were collected and packed into 10 L containers, resulting in four replications for each

  15. Inert matrix fuel in dispersion type fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Savchenko, A.M. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)]. E-mail: sav@bochvar.ru; Vatulin, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Morozov, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Sirotin, V.L. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Dobrikova, I.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kulakov, G.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Ershov, S.A. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kostomarov, V.P. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Stelyuk, Y.I. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)

    2006-06-30

    The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg{sup -1} (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

  16. Inert matrix fuel in dispersion type fuel elements

    Science.gov (United States)

    Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

    2006-06-01

    The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

  17. Gerenciamento de Projetos Segundo o Guia PMBOK: Desafios para os Gestores

    Directory of Open Access Journals (Sweden)

    David Ferreira Bomfin

    2013-01-01

    Full Text Available Analisamos, dentre as dificuldades e facilidades mapeadas na literatura, quais e por que as mesmas estão presentes no processo de gerenciamento de projetos com base no guia PMBOK. Verificamos quais os possíveis argumentos que os gestores vivenciam no gerenciamento de projetos no que diz respeito aos sucessos e fracassos apresentados. O estudo se concentrou em profissionais com certificação PMP por apresentarem o conhecimento em Gerenciamento de Projetos com base no guia PMBOK e, por ser uma área de crescimento onde as empresas buscam profissionais com melhores práticas e vantagem competitiva. Valendo-se do questionário estruturado como instrumento de coleta de dados e analisando os resultados, os principais pontos encontrados dizem respeito ao tempo em que os gerentes lidam com gerenciamento de projetos, as dificuldades encontradas pelos gerentes em sua prática profissional, a área do PMBOK com maior dificuldade de ser gerenciada, os itens que geram sucessos e fracassos nos projetos. A entrevista semi-estruturada evidenciou que o gerenciamento de projetos apresenta algumas dificuldades presentes na prática profissional dos gerentes. Sugerimos investigar melhor o impacto que a certificação PMP proporciona aos gerentes de projetos e se a relação maturidade organizacional e experiência em gerenciamento de projetos é fator de sucesso para o projeto.DOI:10.5585/gep.v3i3.78

  18. Guia de atencion integral para la prevencion, deteccion temprana y tratamiento de las complicaciones del embarazo: Sección de Toxoplasmosis

    Directory of Open Access Journals (Sweden)

    Jorge Alberto Cortes

    Full Text Available Resumen Como parte de la Guia de atencion integral para la prevencion, deteccion temprana y tratamiento de las complicaciones del embarazo, desarrollada y financiada por el Departamento de Ciencia, Tecnologia e Innovacion, Colciencias, y el Ministerio de Salud y Proteccion Social de Colombia, se seleccionó a la infeccion por Toxoplasma para el desarrollo de recomendaciones para su prevencion, diagnostico y tratamiento. La infeccion por Toxoplasma gondii (T. gondii durante el embarazo puede resultar en graves complicaciones para el feto y dejar importantes secuelas al recien nacido. Se realizo una guia basada en la mejor evidencia disponible en la literatura cientifica, con especial pertinencia a la informacion colombiana. Un consenso de expertos en parasitologia, ginecologia, neonatologia e infectologia, tanto de adultos como pediatrica, desarrollo las recomendaciones. Se propone que las recomendaciones de esta guia de atencion integral sean utilizadas por los profesionales de salud de los programas de atencion del embarazo del pais con el fin de disminuir la morbilidad y mortalidad atribuible a esta enfermedad. Se formulan recomendaciones especificas para el diagnostico desde el primer trimestre, consejos de prevencion en las mujeres no infectadas, identificacion de la infeccion del feto o del recien nacido y recomendaciones de tratamiento en estos escenarios.

  19. INTER-RELAÇÃO DA FERRAMENTA APQP E DO GUIA PMBOK PARA EFICIÊNCIA NAS ETAPAS DE IMPLEMENTAÇÃO DE NOVOS PROJETOS NA INDÚSTRIA AUTOMOTIVA

    Directory of Open Access Journals (Sweden)

    Jean Cristhiano Franco

    2014-06-01

    Full Text Available O desenvolvimento de novos projetos na industria automotiva apresenta-se como um projeto de grande porte, regido por metodologias, guias e ferramentas estruturadas que visam garantir a eficácia do projeto. Esta pesquisa foi desenvolvida com o intuito de contribuir para um maior entendimento entre a ferramenta APQP, de grande aplicação na industria automotiva, e o guia PMBOK, considerado referência no gerenciamento de projetos. Neste âmbito, este estudo realiza uma análise comparativa entre suas principais características, identificando suas semelhanças e divergências. Dentre os diversos resultados do estudo, é possível identificar que o guia PMBOK é mais completo que a ferramenta APQP utilizada no setor automobilístico, uma vez que traz de forma mais detalhada as etapas e as técnicas que são indicadas nas atividades, segundo as boas práticas dos PMPs (Project Management Professional.

  20. Dietary recommendations: comparing dietary guidelines from Brazil and the United States Recomendações dietéticas: comparação entre os guias alimentares brasileiro e americano

    Directory of Open Access Journals (Sweden)

    Rosely Sichieri

    2010-11-01

    Full Text Available The Brazilian dietary guidelines are based in part on mainstream United States' recommendations, in spite of the criticisms and shortcomings of the American guidelines. In this paper, Brazilian food guidelines are summarized and discussed in comparison with the USA recommendations. American and Brazilian dietary recommendations are quite similar in many aspects, particularly those related to variety in the diet, the importance of physical activity and weight management. Different to American guidelines, those from Brazil advise people to choose fresh foods, to prefer healthier types of fat, to limit trans fat intake and to eat good sources of protein, but does not recommend the consumption of whole grains. Besides the challenges related to their implementation, indicators for the evaluation of the effectiveness of these guidelines should be established from the beginning, particularly those related to changes in dietary habits and the prevalence of obesity.O guia alimentar brasileiro é baseado parcialmente nas recomendações americanas a despeito das críticas e problemas identificados no documento dos Estados Unidos. Neste artigo, as recomendações alimentares para o Brasil são resumidas e discutidas em comparação com as recomendações estadunidenses. Os guias alimentares brasileiro e americano são bastante similares em diversos aspectos, particularmente aqueles relacionados com a variação da dieta, a importância da atividade física e o gerenciamento do peso. Diferentemente dos Estados Unidos, o guia brasileiro estimula o consumo de alimentos frescos, aconselha o uso de fontes saudáveis de gorduras, a limitação do consumo de gordura trans, o consumo de boas fontes de proteínas, mas não indica o consumo de grãos integrais. Além dos desafios relacionados com a sua implantação, os indicadores para a avaliação da eficácia dos guias alimentares devem ser estabelecidos desde sua implantação, particularmente, os relacionados

  1. Enrichment measurement in TRIGA type fuels

    International Nuclear Information System (INIS)

    Aguilar H, F.; Mazon R, R.

    2001-05-01

    The Department of Energy of the United States of North America, through the program 'Idaho Operations Nuclear Spent Fuel Program' of the Idaho National Engineering and Environmental Laboratory (INEEL), in Idaho Falls; Idaho USA, hires to Global Technologies Inc. (GTI) to develop a prototype device of detection enrichment uranium (DEU Detection of Enrichment of Uranium) to determine quantitatively the enrichment in remainder U-235 in a TRIGA fuel element at the end of it useful life. The characteristics of the prototype developed by GTI are the following ones: It allows to carry out no-destructive measurements of TRIGA type fuel. Easily transportable due to that reduced of it size. The determination of the enrichment (in grams of U-235) it is obtained with a precision of 5%. The National Institute of Nuclear Research (ININ), in its facilities of the Nuclear Center of Mexico, it has TRIGA type fuel of high and low enrichment (standard and FLIP) fresh and with burnt, it also has the infrastructure (hot cells, armor-plating of transport, etc) and qualified personnel to carry out the necessary maneuvers to prove the operation of the DEU prototype. For this its would be used standard type fuel elements and FLIP, so much fresh as with certain burnt one. In the case of the fresh fuels the measurement doesn't represent any risk, the fuels before and after the measurement its don't contain a quantity of fission products that its represent a radiological risk in its manipulation; but in the case of the fuels with burnt the handling of the same ones represents an important radiological risk reason why for its manipulation it was used the transport armor-plating and the hot cells. (Author)

  2. Fuel assembly for FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki.

    1995-01-01

    Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

  3. Ver, Compreender e Aprender: estudo comparado com guias de uso de apps editoriais para dispositivos móveis

    Directory of Open Access Journals (Sweden)

    Ana Claudia Dalagnoli

    2016-12-01

    Full Text Available Este estudo foi construído a partir da percepção de que a evolução das tecnologias e, sobretudo, o aprimoramento da comunicação em dispositivos móveis, impacta diretamente no processo de desenvolvimento de aplicativos (apps e softwares e na relação do ser humano com as novas interfaces. A pesquisa apresenta uma análise comparativa entre os aplicativos para iPad das revistas de decoração e design de interiores Casa Claudia (Editora Abril e Casa e Jardim (Editora Globo, a partir de testes de usabilidade dos mesmos em tablets Apple. O objetivo foi verificar se os guias de uso disponíveis em aplicativos (apps de revistas para dispositivos móveis, construídos por textos e imagens, estão cumprindo a função de comunicar, efetivamente, o modo de uso desses apps. As informações foram coletadas junto a mulheres com idade entre 25 e 54 anos que testaram e comentaram o funcionamento das versões atuais dos apps. Foram selecionadas 24 mulheres para fazerem atividades individuais monitoradas em ambos os apps, com dados coletados por meio de observações e think aloud. Os dados obtidos foram analisados, e com base nos resultados foi proposto um ajuste nos requisitos para o desenvolvimento de interfaces de guias para aplicativos com vistas à maior satisfação das usuárias. Foram percebidas falhas severas na apreensão e aplicação das orientações disponíveis nos guias de uso. Os pesquisadores verificaram que os requisitos básicos para aumentar a satisfação e melhorar a usabilidade estão relacionados à facilidade de compreensão e à simplicidade no uso, o que pode ser obtido por meio de melhorias relacionadas à aplicação da tecnologia, ao conhecimento dos processos cognitivos dos usuários e à aplicação de conceitos semióticos de comunicação. O estudo pode servir de base para o desenvolvimento e análise de guias de uso de outros apps com características similares, em diversos sistemas operacionais.

  4. Fuel exchanger in FBR type reactor

    International Nuclear Information System (INIS)

    Shinden, Kazuhiko; Tanaka, Osamu.

    1990-01-01

    The present invention concerns a fuel exchanger for exchanging fuels in an LMFBR type reactor using liquid metals as coolants. An outer gripper cylinder rotating device for rotating an outer gripper cylinder that holds a gripper is driven, to lower the gripper driving portion and the outer gripper cylinder, fuels are caught by the finger at the top end of the outer gripper cylinder and elevated to extract the fuels from the reactor core. Then, the gripper driving portion casing and the outer gripper cylinder are rotated to rotate the fuels caught by the gripper. Subsequently, the gripper driving portion and the outer gripper cylinder are lowered to charge the fuels in the reactor core. This can directly shuffle the fuels in the reactor core without once transferring the fuels into a reactor storing pot and replacing with other fuels, thereby shortening the shuffling time. (I.N.)

  5. Paired replacement fuel assemblies for BWR-type reactor

    International Nuclear Information System (INIS)

    Oguchi, Kazushige.

    1997-01-01

    There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

  6. Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

    International Nuclear Information System (INIS)

    Delafosse, Jacques.

    1976-01-01

    French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

  7. CANDU type fuel behavior evaluation - a probabilistic approach

    International Nuclear Information System (INIS)

    Moscalu, D.R.; Horhoianu, G.; Popescu, I.A.; Olteanu, G.

    1995-01-01

    In order to realistically assess the behavior of the fuel elements during in-reactor operation, probabilistic methods have recently been introduced in the analysis of fuel performance. The present paper summarizes the achievements in this field at the Institute for Nuclear Research (INR), pointing out some advantages of the utilized method in the evaluation of CANDU type fuel behavior in steady state conditions. The Response Surface Method (RSM) has been selected for the investigation of the effects of the variability in fuel element computer code inputs on the code outputs (fuel element performance parameters). A new developed version of the probabilistic code APMESRA based on RSM is briefly presented. The examples of application include the analysis of the results of an in-reactor fuel element experiment and the investigation of the calculated performance parameter distribution for a new CANDU type extended burnup fuel element design. (author)

  8. O simbolismo da águia na religiosidade nórdica pré-cristã e cristã

    Directory of Open Access Journals (Sweden)

    Johnni Langer

    2015-10-01

    Full Text Available Presente desde a Pré-História até o fim da Idade Média, as representações de águias no mundo nórdico assumiram diversos simbolismos religiosos. Em busca de compreender as diferentes funções deste animal em diferentes períodos, abordaremos o simbolismo da águia de forma distinta nos períodos Pré-Viking, na Era Viking e no Cristianismo da Escandinávia Medieval. Para tanto, realizaremos uma análise de diversas fontes iconográficas, arqueológicas e literárias, como jóias, monumentos e manuscritos, em um diálogo com autores clássicos — Hilda Davidson, James Graham-Campbell e Régis Boyer, por exemplo —, e com estudos atualizados de especialistas no assunto, tais como Anne-Sophie Gräslund, Jens Peter Schjødt e Kristina Jennbert. Como referenciais teóricos e metodológicos, adotamos o conceito de longa duração na abordagem de Lotte Hedeager e o conceito de símbolo para a Arqueologia das Religiões, aplicados à religiosidade nórdica.

  9. Effect of aviation fuel type and fuel injection conditions on the spray characteristics of pressure swirl and hybrid air blast fuel injectors

    Science.gov (United States)

    Feddema, Rick

    Feddema, Rick T. M.S.M.E., Purdue University, December 2013. Effect of Aviation Fuel Type and Fuel Injection Conditions on the Spray Characteristics of Pressure Swirl and Hybrid Air Blast Fuel Injectors. Major Professor: Dr. Paul E. Sojka, School of Mechanical Engineering Spray performance of pressure swirl and hybrid air blast fuel injectors are central to combustion stability, combustor heat management, and pollutant formation in aviation gas turbine engines. Next generation aviation gas turbine engines will optimize spray atomization characteristics of the fuel injector in order to achieve engine efficiency and emissions requirements. Fuel injector spray atomization performance is affected by the type of fuel injector, fuel liquid properties, fuel injection pressure, fuel injection temperature, and ambient pressure. Performance of pressure swirl atomizer and hybrid air blast nozzle type fuel injectors are compared in this study. Aviation jet fuels, JP-8, Jet A, JP-5, and JP-10 and their effect on fuel injector performance is investigated. Fuel injector set conditions involving fuel injector pressure, fuel temperature and ambient pressure are varied in order to compare each fuel type. One objective of this thesis is to contribute spray patternation measurements to the body of existing drop size data in the literature. Fuel droplet size tends to increase with decreasing fuel injection pressure, decreasing fuel injection temperature and increasing ambient injection pressure. The differences between fuel types at particular set conditions occur due to differences in liquid properties between fuels. Liquid viscosity and surface tension are identified to be fuel-specific properties that affect the drop size of the fuel. An open aspect of current research that this paper addresses is how much the type of aviation jet fuel affects spray atomization characteristics. Conventional aviation fuel specifications are becoming more important with new interest in alternative

  10. Effects of Fuel Type and Fuel Delivery System on Pollutant Emissions of Pride and Samand Vehicles

    Directory of Open Access Journals (Sweden)

    Akbar Sarhadi

    2017-04-01

    Full Text Available This research was aimed to study the effect of the type of fuel delivery system (petrol, dedicated or bifuel, the type of consumed fuel (petrol or gas, the portion of consumed fuel and also the duration of dual-fuelling in producing carbon monoxide, carbon dioxide and unburned hydrocarbons from Pride and Samand. According to research objectives, data gathering from 2000 vehicles has been done by visiting Hafiz Vehicle Inspection Center every day for 2 months. The results of this survey indicated that although there is no significant difference between various fuel delivery systems in terms of producing the carbon monoxide, carbon dioxide and unburned hydrocarbons by Samand, considering the emission amount of carbon dioxide, the engine performance of Pride in bifuel and dedicated state in GTXI and 132 types is more unsatisfactory than that of petrol state by 0.3 and 0.4%, respectively. On the other hand, consuming natural gas increases the amount of carbon monoxide emission in dual- fuel Pride by 0.18% and decreases that in dual-fuel Samand by 1.2%, which signifies the better design of Samand in terms of fuel pumps, used kit type and other engine parts to use this alternative fuel compared to Pride. Since the portion of consumed fuel and also duration of dual-fuelling does not have a significant effect on the amount of output pollutants from the studied vehicles, it can be claimed that the output substances from the vehicle exhaust are more related to the vehicle’s condition than the fuel type.

  11. Guia per al càlcul simplificat de corbes de valoració

    OpenAIRE

    Núñez Burcio, Oscar; Serrano i Plana, Núria; Subirats i Vila, Xavier

    2016-01-01

    La guia descriu com calcular corbes de valoració d'acids i bases, utilitzant simplificacions, a partir de la seva concentració i de les constants de dissociació. Els casos presentats són els següents: - Valoració d’un àcid fort amb una base forta - Valoració d’una base forta amb un àcid fort - Valoració d’un àcid feble monopròtic amb una base forta - Valoració d’una base feble monopròtica amb un àcid fort - Valoració d’un àcid feble dipròtic amb una base forta - Valoració d...

  12. Fuel type characterization based on coarse resolution MODIS satellite data

    Directory of Open Access Journals (Sweden)

    Lanorte A

    2007-01-01

    Full Text Available Fuel types is one of the most important factors that should be taken into consideration for computing spatial fire hazard and risk and simulating fire growth and intensity across a landscape. In the present study, forest fuel mapping is considered from a remote sensing perspective. The purpose is to delineate forest types by exploring the use of coarse resolution satellite remote sensing MODIS imagery. In order to ascertain how well MODIS data can provide an exhaustive classification of fuel properties a sample area characterized by mixed vegetation covers and complex topography was analysed. The study area is located in the South of Italy. Fieldwork fuel type recognitions, performed before, after and during the acquisition of remote sensing MODIS data, were used as ground-truth dataset to assess the obtained results. The method comprised the following three steps: (I adaptation of Prometheus fuel types for obtaining a standardization system useful for remotely sensed classification of fuel types and properties in the considered Mediterranean ecosystems; (II model construction for the spectral characterization and mapping of fuel types based on two different approach, maximum likelihood (ML classification algorithm and spectral Mixture Analysis (MTMF; (III accuracy assessment for the performance evaluation based on the comparison of MODIS-based results with ground-truth. Results from our analyses showed that the use of remotely sensed MODIS data provided a valuable characterization and mapping of fuel types being that the achieved classification accuracy was higher than 73% for ML classifier and higher than 83% for MTMF.

  13. Fueling method in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Inoue, Kotaro.

    1985-01-01

    Purpose: To extend the burning cycle and decrease the number of fuel exchange batches without increasing the excess reactivity at the initial stage of burning cycles upon fuel loading to an LMFBR type reactor. Method: Each of the burning cycles is divided into a plurality of burning sections. Fuels are charged at the first burning section in each of the cycles such that driver fuel assemblies and blanket assemblies or those assemblies containing neutron absorbers such as boron are distributed in mixture in the reactor core region. At the final stage of the first burning section, the blanket assemblies or neutron absorber-containing assemblies present in mixture are partially or entirely replaced with driver fuel assemblies depending on the number of burning sections such that all of them are replaced with the driver fuel assemblies till the start of the final burning section of the abovementioned cycle. The object of this invention can thus be attained. (Horiuchi, T.)

  14. Technical report: fabrication of PWR type rodlet fuel

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Uno, Hisao; Sasajima, Hideo

    1990-06-01

    With respect to the simulated reactivity initiated accident (RIA) experiments with pre-irradiated LWR type fuel rods at nuclear safety research reactor (NSRR), there were principally three technical difficulties which should be overcome: (1) Fabrication of the rodlet fuel; Fuel rods from the commercial power reactors had an active column length by 3.6m. To utilize this for NSRR pulse experiment, rodlet fuel having an active column length by 0.12m (reduced to one thirtieth) is requested to fabricate without changing the inside fuel conditions. (2) Development of in-core instrumentations: During pre-irradiation stages, a long-sized fuel rod had dimensional changes by waterside corrosion, bowing, creep down and so on. The fuel also had greater amount of radioactive fission products. This condition is significant to in-core instrumentations to be attached to the fuel rods. Well characterized data to be obtained from these, however, are quite necessary and important from research point of view. Remote handling techniques to attach the rod pressure sensor, the cladding extensometer, the fuel extensometer, and the cladding surface thermocouple to pre-irradiated fuel rods are, therefore, requested to develop. (3) Installation of PIE equipments for pulsed rodlet fuels: PIE on the pulsed rodlet fuels are necessary to better understanding the fuel performance detaily. Equipments which can easily detect the data related to PCMI type fuel failure are matter of concern. Since 1986, the technical difficulties have been tried to overcome by all staffs belonging to Reactivity Accident Laboratory, NSRR Operation Division, Department of Reactor Fuel Examination and Hot Laboratory. This report describes the technical achievements obtained through four years work. (author)

  15. The feasibility study on fuel types for the KALIMER

    International Nuclear Information System (INIS)

    Hwang, W.; Nam, C.; Yim, J. S.; Na, B. C.; Hahn, D. H.; Kim, Y. I.; Kim, Y. C.; Park, C. K.

    1997-08-01

    The economics of LMR is largely dependent on the construction cost of the power plant, and the fuel cycle options usually constitute 20 to 30 % of total electricity generation cost. The choice of fuel cycle technology and the fuel type is important in order to develop a LMR with better economics, performance and safety. The LMR fuel types, whose performances have been proven up to 15 at% burnup, are MOX and IFR metal fuel. The base alloy, binary (U-10% Zr) metal fuel with HT9 is used as structural materials of KALIMER. The design concept of KALIMER fuel has been established through the investigation of technical feasibilities on the fuel and recycle systems for MOX and IFR metal fuel. According to the results of comparative analysis for MOX and metal fuel, metal fuel is better than MOX in view of safety, in-reactor performance, nuclear characteristics, economics and non-proliferation, while MOX fuels have advantages in the developmental status and technical cooperation potential. The overall performance of binary (U-10% Zr) metal fuel with HT9 cladding, which is a potential start-up fuel for KALIMER, is not only superior to that of MOX fuel, but also has enough technical feasibility in its high-burnup performance, safety and economics. (author). 54 ref., 13 tabs., 20 figs

  16. O encontro entre o desenvolvimento rural sustentável e a promoção da saúde no Guia Alimentar para a População Brasileira

    OpenAIRE

    Oliveira, Nádia Rosana Fernandes de; Jaime, Patricia Constante

    2016-01-01

    Resumo O Guia Alimentar é documento que aborda os princípios e as recomendações de uma alimentação adequada e saudável para a população brasileira, tendo como propósito apoiar a educação alimentar e nutricional e subsidiar políticas e programas nacionais de alimentação e nutrição no setor de saúde e também em outros setores. O objetivo deste estudo foi identificar a interseção entre a promoção da saúde e o desenvolvimento rural sustentável no Guia Alimentar para a População Brasileira. Realiz...

  17. 14 CFR 26.37 - Pending type certification projects: Fuel tank flammability.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Pending type certification projects: Fuel tank flammability. 26.37 Section 26.37 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION... AIRPLANES Fuel Tank Flammability § 26.37 Pending type certification projects: Fuel tank flammability. (a...

  18. Evaluation of plate type fuel options for small power reactors

    International Nuclear Information System (INIS)

    Andrzejewski, Claudio de Sa

    2005-01-01

    Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO 2 in stainless steel, of UO 2 in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

  19. Sabores de viagem - as culinárias regionais nos guias turísticos entre 1966 e 1983.

    Directory of Open Access Journals (Sweden)

    Luciana Patrícia de Morais

    2012-11-01

    Full Text Available Partindo da premissa de que os saberes culinários se constituem em um dos principais suportes nos processos de construção e transmissão de identidades, busca-se entender a comida regional como expressão das culturas regionais em sua relação com a atividade turística. Parte-se do pressuposto de que a difusão dos regionalismos tem relação próxima com o desenvolvimento desta atividade no Brasil, num movimento em que as personalidades regionais são apropriadas e alçadas ao status de atrativos e produtos turísticos. Sem pretender estabelecer uma relação de causalidade entre turismo e comida regional, entende-se que embora a visibilidade das culinárias típicas no Brasil tenha relação com uma série de outras variáveis, o turismo é fundamental para se compreender os contornos e o alcance da visibilidade destas culinárias. Para tanto, a discussão terá como ponto de partida uma reflexão sucinta sobre os guias de viagem como fonte de pesquisa tomando como estudo de caso específico os casos de Minas Gerais e Paraná no interior do Guia Quatro Rodas Brasil entre 1966 e 1983.

  20. R and D activities on CANDU-type fuel in Indonesia

    International Nuclear Information System (INIS)

    Suripto, A.; Badruzzaman, M.; Latief, A.

    1997-01-01

    The status of R and D activities in Indonesia with respect of CANDU-type fuel development is presented. The activities have been started since the first feasibility study to introduce nuclear power plants was carried out in 1970s. The early research comprised the in-situ pilot production of yellow-cake in Kalimantan (Borneo) experimental mining site, uranium purification and pellet preparation. This program continued to gain a full support from the Government which culminated in the realisation of the construction by BATAN of a large fuel development laboratory in Serpong, starting from 1984 in co-operation with NIRA Ansaldo of Italy. The laboratory, which is called the Power Reactor Experimental Fuel Element Installation (EFEI) was originally designed as an experimental facility to integrate the acquired domestic R and D results gained so far on the CANDU-type fuel technology and the additional know-how received from NIRA Ansaldo which at that time was engaged, in developing a CANDU-type fuel, called the CIRENE fuel design. In the present days the facility houses the power reactor fuel development activities carried out to build up the national capability on power reactor fuel fabrication technology in anticipation to embark upon the nuclear energy era in the near future. (author)

  1. Fuel assembly for FBR type reactor and reactor core thereof

    International Nuclear Information System (INIS)

    Kobayashi, Kaoru.

    1998-01-01

    The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

  2. 14 CFR 26.33 - Holders of type certificates: Fuel tank flammability.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Holders of type certificates: Fuel tank... Tank Flammability § 26.33 Holders of type certificates: Fuel tank flammability. (a) Applicability. This... part 25 of this chapter. (2) Exception. This paragraph (b) does not apply to— (i) Fuel tanks for which...

  3. Adesao ao guia alimentar para populacao brasileira

    Directory of Open Access Journals (Sweden)

    Eliseu Verly Junior

    2013-12-01

    Full Text Available OBJETIVO : Analisar a adesão ao Guia Alimentar para População Brasileira. MÉTODOS : Amostra composta por participantes do Inquérito de Saúde de São Paulo (n = 1.661 que preencheram dois recordatórios de 24 horas. Foi utilizado modelo bivariado de efeito misto para a razão entre o consumo de energia do grupo de alimentos e o consumo calórico total. A razão estimada foi utilizada para calcular o percentual de indivíduos com consumo abaixo ou acima da recomendação. RESULTADOS : Pelo menos 80,0% da população consome abaixo do recomendado para: leite e derivados; frutas e sucos de frutas; e cereais, tubérculos e raízes; aproximadamente 60,0% para legumes e verduras; 30,0% para feijões; e 8,0% para carnes e ovos. Adolescentes apresentaram a maior inadequação para legumes e verduras (90,0%, e o estrato de maior renda foi associado à menor inadequação para óleos, gorduras e sementes oleaginosas (57,0%. CONCLUSÕES : Foi observado consumo inadequado dos grupos de alimentos relacionados com aumento do risco de doenças crônicas.

  4. Measurement of the Velocity and Pressure Drop in a Tubular Type Fuel

    International Nuclear Information System (INIS)

    Jonghark Park; Heetaek Chae; Cheol Park; Heonil Kim

    2006-01-01

    We have developed a tubular type fuel assembly design as one of candidates for fuel to be used in the Advanced HANARO Reactor (AHR). The tubular type fuel has several merits over a rod type fuel with respect to the thermal-hydraulic and structural safety; the larger ratio of surface area to volume makes the surface temperature of a fuel element become lower, and curved plate is stronger against longitudinal bending and vibration. In the other side, a disadvantage is expected such that the flow velocity can be distributed unevenly channel by channel because the flow channels are isolated from each other in a tubular type fuel assembly. In addition to the design development, we also investigated the flow characteristics of the tubular fuel experimentally. To examine the flow velocity distribution and pressure drop, we made an experiment facility and a mockup of the tubular fuel assembly. The fuel assembly consists of 6 concentric fuel tubes so that 7 layers are made between fuel tubes. Since each layer is divided into three sections by stiffeners, 21 isolated flow channels are made in total. We employed pitot-tubes to measure the coolant velocity in each channel. The maximum velocity was measured as large as about 28% of the average velocity. It was observed in the innermost channel contrarily to the expectation from the hydraulic diameter. A change in the total flow rate did not affect the flow distribution. Meanwhile, the pressure drop was measured as about 70% of the drop in the rod type fuel assembly in use in HANARO. (authors)

  5. Analyses for licensing of new fuel types at Paks NPP

    International Nuclear Information System (INIS)

    Kereszturi, A.; Bogatyr, S.; Miko, S.; Nemes, I.

    2003-01-01

    In the last years Paks NPP initiated several projects aiming at the introduction of new fuel types and resulting in more economic fuel cycles. The motivations, the reasons, and the economic consequences of the above modifications are detailed. The application of a new fuel type requires the renewal of the relevant chapters of the Safety Analysis Report. The fulfilment of fuel design basis requirements, to be summarised briefly also in the paper, must be investigated during normal and accidental conditions. The characteristics of the different codes, the data transfer between them are detailed. After, the cases of the Normal Operation, Anticipated Operation Occurrence, and the Postulated Accidents, judged as the most relevant ones in case of fuel modifications, are overviewed. In the last part, selected examples of the licensing calculations, performed by the above tools are presented. In conclusion, modifications of the WWER fuel, namely increased enrichment, application of burnable fuel pins, modified geometry make more economic fuel cycles (larger discharge burnup, power up-rate, reduced pressure vessel fluence) are possible. The further step (increased enrichment, burnable poison) of the fuel modernisation at NPP Paks is necessary for more economic fuel cycles and fuel consuming. A sound basis of licensing methodology, safety analysis, and necessary computer codes for the WWER fuel modernisation is available

  6. Fabrication and characterization of MX-type fuels and fuel pins

    International Nuclear Information System (INIS)

    Richter, K.; Bartscher, W.; Benedict, U.; Gueugnon, J.F.; Kutter, H.; Sari, C.; Schmidt, H.E.

    1978-01-01

    This paper summarizes the most important fabrication parameters and characterization of fuel and fuel pins obtained during the investigation of uranium-plutonium carbides, oxicarbides, carbonitrides and nitrides in the past years at the European Institute for Transuranium Elements at Karlsruhe. All preparation methods discussed are based on carbothermic reduction of a mechanical blend of uranium-plutonium oxide and carbon powder. General data for carbothermic reduction processes are discussed (influence of starting material, homogeneity, control of degree of reaction, etc). A survey of different preparation methods investigated is given. Limitations with respect to temperature and atmosphere for both carbothermic reduction processes and sintering conditions for the different compounds are summarized. A special preparation process for mixed carbonitrides with low nitrogen content (U,Pu)sub(1-x)Nsub(x) in the range 0.1 0 C to 1400 0 C by means of a modulated electron beam technique. A scheme is proposed, which allows to predict the thermal properties of MX fuels on the basis of their chemical composition and porosity. Preparation, preirradiation characterization and final controls of fuel test pins for pellet and vibrocompacted type of pins are described and the most important data summarized for all advanced fuels irradiated at Dounreay (DN1) and Rapsodie Fast Reactor (DN2) within the TU irradiation programme

  7. Development of core technology for research reactors using plate type fuels

    International Nuclear Information System (INIS)

    Ha, Jae Joo; Lee, Doo Jeong; Park, Cheol

    2009-12-01

    Around 250 research reactors are under operation over the world. However, about 2/3 have been operated more than 30 years and demands for replacements are expected in the near future. The number of expected units is around 110, and around 55 units from 40 countries will be expected to be bid in the world market. In 2007, Netherlands started international bidding process to construct a new 80MW RR (named PALLAS) with the target of commercial operation in 2016, which will replace the existing HFR(45MW). KAERI consortium has been participated in that bid. Most of RRs use plate type fuels as a fuel assembly, Be and Graphite as a reflector. On the other hand, in Korea, the KAERI is operating the HANARO, which uses a rod type fuel assembly and heavy water as a reflector. Hence, core technologies for RRs using plate type fuels are in short. Therefore, core technologies should be secured for exporting a RR. In chapter 2, the conceptual design of PALLAS which use plate type fuels are described including core, cooling system and connected systems, layout of general components. Experimental verification tests for the plate type fuel and second shutdown system and the code verification for nuclear design are explained in Chapter 3 and 4, respectively

  8. Fuel assemblies for BWR type reactors

    International Nuclear Information System (INIS)

    Ishizuka, Takao.

    1981-01-01

    Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

  9. Design characteristics of pantograph type in vessel fuel handling system in SFR

    International Nuclear Information System (INIS)

    Kim, S. H.; Koo, G. H.

    2012-01-01

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied

  10. Design characteristics of pantograph type in vessel fuel handling system in SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. H.; Koo, G. H. [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied.

  11. Stressed and strained state for cermetic-rod-type fuel element

    International Nuclear Information System (INIS)

    Kulikov, I.S.

    1987-01-01

    Calculation technique for designing the stress-strained state of a cermetic rod-type fuel element has been proposed. The technique is based on the time-dependent step-by-step method and the solution of the deformation equilibrium equation for continuous and thick-wall long cylinders at every temporal step by the finite difference method. Additional strains, caused by thermal expansion and radiation swelling, have been taken into account. The transion from the non-contact model to the stiff-contact model has been provided in the case of cladding-fuel gap dissappearing in one or a number of cross-sections along the fuel element height. The method is supplemented by the formula for fuel cans stability estimation in the case of high coolant external pressure. The example of estimation of the cermetic-rod-type fuel elements are considered as an example

  12. Fuel performance of rod-type research reactor fuel using a centrifugally atomized U-Mo powder

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Lee, Yoon Sang; Kim, Chang Kyu

    2009-01-01

    A low enriched uranium nuclear fuel for research reactors has been developed in order to replace a highly enriched uranium fuel according to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program. In KAERI, a rod-type U 3 Si dispersion fuel has been developed for a localization of the HANARO fuel and a U 3 Si/Al dispersion fuel of 3.15 gU/cc has been used at HANARO as a driver fuel since 2005. Although uranium silicide dispersion fuels such as U 3 Si 2 /Al and U 3 Si/Al are being used widely, high uranium density dispersion fuels (8-9 g/cm 3 ) are required for some high performance research reactors. U-Mo alloys have been considered as one of the most promising uranium alloys for a dispersion fuel due to their good irradiation performance. An international qualification program on U-Mo fuel to replace a uranium silicide dispersion fuel with a U-Mo dispersion fuel has been carried out

  13. The prospects of use of alternative types of fuel in road transport ...

    African Journals Online (AJOL)

    The article is devoted to the analysis of possibilities of using alternative types of fuel in transport. Gas engine fuels are considered as potential energy carriers for diesel engines. Since the constructions of vehicles, using gas and traditional types of fuel, have some differences, the most important are the issues of ensuring ...

  14. In-use vs. type-approval fuel consumption of current passenger cars in Europe

    International Nuclear Information System (INIS)

    Ntziachristos, L.; Mellios, G.; Tsokolis, D.; Keller, M.; Hausberger, S.; Ligterink, N.E.; Dilara, P.

    2014-01-01

    In-use fuel consumption data of 924 passenger cars (611 petrol, 313 diesel) were collected from various European sources and were evaluated in comparison to their corresponding type-approval values. The analysis indicated that the average in-use fuel consumption was higher than the type-approval one by 11% for petrol cars and 16% for diesel cars. Comparison of this dataset with the Travelcard database in the Netherlands showed that the deviation increased for late model years and in particular for cars with low type-approval values. The deviation was higher than 60% for vehicles registered in 2012 within the 90–100 gCO 2 /km bin. Unrealistic vehicle resistances used in type-approval were identified as one of the prime reasons of the difference. A simplified linear model developed in the study may be used to predict in-use fuel consumption based on data publicly available. The model utilizes the fuel consumption measured in type-approval, the mass, and the engine capacity to provide in-use fuel consumption. This may be either used to correct fuel consumption factors currently utilized by emission models (e.g. COPERT, HBEFA, VERSIT+, and others) or could be used independently to make projections on how fuel consumption may develop on the basis of changing future passenger cars characteristics. - Highlights: • In-use fuel consumption of petrol and diesel passenger cars is 11% and 16% higher than type-approval, respectively. • The relative difference between in-use and type-approval increases for late model and vehicles with low consumption. • Unrealistically low vehicle resistances are identified as a prime reason of low type-approval fuel consumption. • A model developed predicts in-use consumption on the basis of type-approval consumption, vehicle mass, and engine capacity

  15. Control in fabrication of PWR and BWR type reactor fuel elements

    International Nuclear Information System (INIS)

    Gorskij, V.V.

    1981-01-01

    Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

  16. A CAREM type fuel element dynamic analysis

    International Nuclear Information System (INIS)

    Magoia, J.E.

    1990-01-01

    A first analysis on the dynamic behaviour of a fuel element designed for the CAREM nuclear reactor (Central Argentina de Elementos Modulares) was performed. The model used to represent this dynamic behaviour was satisfactorily evaluated. Using primary estimations for some of its numerical parameters, a first approximation to its natural vibrational modes was obtained. Results obtained from fuel elements frequently used in nuclear power plants of the PWR (Pressurized Water Reactors) type, are compared with values resulting from similar analysis. (Author) [es

  17. Electromagnetic Acoustic Test of the Artificial Defects for a Plate-type Nuclear Fuel

    International Nuclear Information System (INIS)

    Jung, Hyun Kyu; Kim, Dong Min; Lee, Yoon Sang; Cheong, Yong Moo

    2011-01-01

    Most research and test reactors use the nuclear fuel plates which are consisted of a fuel meat in aluminum alloy. Last year, KAERI signed a deal with the Jordan Atomic Energy Commission to build the research reactor and have to supply the plate-type nuclear fuels. For the demands of world market, KAERI started the research and development of the plate-type fuel elements and endeavored to achieve a localization of the plate-type fuel fabrication. For the inspection of plate-type fuel elements to be used in Research Reactors, an immersion pulse-echo ultrasonic technique was applied. This inspection was done under immersion condition, so a nuclear fuel was immersed to be prone to corrosion and needed to have time and cost due to an additional process. The sample that will be examined is a non-ferromagnetic material such as aluminum with a good acousto-elastic property, which requires an effective inspection of a bond quality for a nuclear fuel under a manufacturing environment. The purpose of this study is to investigate the feasibility of an Electromagnetic Acoustic Transducer (EMAT) technology for an automated inspection of a nuclear fuel without water

  18. Número e espaçamento entre hastes de guia de onda para medida da umidade do solo com TDR Number and spacing between wave guide rods for measurement of soil water content with TDR

    Directory of Open Access Journals (Sweden)

    Eugênio F. Coelho

    2003-08-01

    Full Text Available Com este trabalho, objetivou-se avaliar o desempenho de diferentes modelos matemáticos para ajuste dos dados de umidade, em função da constante dielétrica aparente do solo, e a viabilidade de uso de guias de onda de duas e três hastes, com diferentes espaçamentos. Amostras de solo deformadas foram acondicionadas em segmentos de tubos de PVC de 0,075 m de diâmetro. Construíram-se 24 guias de onda com capacitor e 24 guias de onda sem capacitor, sendo que, para cada tipo, 12 guias de onda de duas hastes e 12 de três hastes, com espaçamento entre hastes de 0,009 a 0,022 m, de impedâncias previamente determinadas, foram inseridas no solo, após a saturação em cada segmento de tubo. Dados de umidade do solo determinados por gravimetria e da constante dielétrica aparente obtidos pelo analisador Trase System, foram tomados em cada recipiente durante a secagem do solo de 0,31 a 0,13 m³ m-3. Cinco modelos matemáticos foram ajustados aos dados de umidade em função da correspondente constante dielétrica aparente do solo e o efeito dos modelos na obtenção da umidade do solo, em função da constante dielétrica aparente, foi avaliado estatisticamente. O modelo de Malicki foi o de melhor ajuste dos dados da constante dielétrica aparente, em função da umidade do solo. As guias de onda de três hastes, de diâmetro 0,003 m, comprimento 0,15 m, com espaçamentos entre hastes de 0,017 m, foram as de melhor desempenho, enquanto as guias de onda de três hastes sem capacitor indicaram melhor desempenho na determinação da umidade do solo, comparadas com as de duas hastes sem capacitor. As guias de onda de três hastes sem capacitor apresentaram melhor desempenho na determinação da umidade do solo que as de três hastes com capacitor.The objective of this study was to evaluate the performance of different mathematical models to adjust the humidity data as a function of the apparent dielectric constant of the soil and to evaluate the

  19. Fast reactor fuel pin behavior analyses in a LOF type transient event

    International Nuclear Information System (INIS)

    Mizuno, Tomoyasu; Koyama, Shin-ichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

    2013-06-01

    In order to evaluate integrity limiting parameters of fuel pins during fast reactor core transient events, such as fuel center line temperature and cladding maximum temperature, fuel pin behavior calculations were made using the fast reactor fuel pin performance code CEDAR. The temperature histories of fuel pins during a loss of flow (LOF) type transient events was calculated based on Ross and Stoute type gap conductance model and constant gap conductance model, which is used in a core transient calculation code like HIPRAC. The calculated maximum temperatures of cladding and adjacent coolant channel were lower in the case with Ross and Stoute type model than in the case of constant gap conductance model due to the dynamic change of gap conductance of former case. It is indicated that core transient calculations with constant gap conductance give conservative cladding and coolant temperatures than that with Ross and Stoute type gap conductance model which is thought to be realistic. (author)

  20. Fuel assemblies for use in BWR type reactors

    International Nuclear Information System (INIS)

    Hirukawa, Koji.

    1987-01-01

    Purpose: To moderate the peak configuration of the burnup degree change curve for the infinite multiplication factor by applying an improvement to the arrangement of fuel rods. Constitution: In a fuel assembly for a BWR type reactor comprising a plurality of fuel rods and water rods arranged in a square lattice, fuel rods containing burnable poisons are arranged at four corners at the second and the third layers from the outside of the square lattice arrangement. Among them, the Cd poison effect in the burnable poison incorporated fuel rods disposed at the second layer is somewhat greater at the initial burning stage and then rapidly decreased along with burning. While on the other hand, the poison effect of the burnable poison-incorporated fuel rods at the third layer is smaller than that at the second layer at the initial burning stage and the reduction in the poison effect due to burning is somewhat more moderate. Since these fuel rods are in adjacent with each other, they interfere to each other and also provide an effect of moderating the burning of the burnable poisons. (Takahashi, M.)

  1. Um guia de recomendações para o emprego de dados multimídia em sites web

    OpenAIRE

    Correa, Rodrigo Stéfani

    2002-01-01

    Dissertação (mestrado) - Universidade Federal de Santa Catarina, Centro Tecnológico. Programa de Pós-Graduação em Ciência da Computação. Com a evolução do suporte tecnológico da Web/Internet, surgiram diversos sites web multimídia. Atualmente existem poucos trabalhos voltados à orientação dos desenvolvedores quanto ao uso adequado dos dados multimídia. Este trabalho propõe um guia de recomendações para o emprego de dados multimídia para auxiliar o projeto de sites web. Para alcançar os obj...

  2. Effect of engine parameters and gaseous fuel type on the cyclic variability of dual fuel engines

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed Y.E. Selim [United Arab Emirates University, Al-Ain (United Arab Emirates). Mechanical Engineering Department, Faculty of Engineering

    2005-05-01

    This paper presents an analysis of the cycle-to-cycle combustion variation as reflected in the combustion pressure data of a single cylinder, naturally aspirated, four stroke, Ricardo E6 engine converted to run as dual fuel engine on diesel and gaseous fuel of LPG or methane. A measuring set-up consisting of a piezo-electric pressure transducer with charge amplifier and fast data acquisition card installed on an IBM microcomputer was used to gather the data of up to 1200 consecutive combustion cycles of the cylinder under various combination of engine operating and design parameters. These parameters included type of gaseous fuel, engine load, compression ratio, pilot fuel injection timing, pilot fuel mass, and engine speed. The data for each operating conditions were analyzed for the maximum pressure, the maximum rate of pressure rise representing the combustion noise, and indicated mean effective pressure. The cycle-to-cycle variation is expressed as the mean value, standard deviation, and coefficient of variation of these three parameters. It was found that the type of gaseous fuel and engine operating and design parameters affected the combustion noise and its cyclic variation and these effects have been presented. 21 refs., 6 figs., 1 tab.

  3. Power distribution gradients in WWER type cores and fuel failure root causes

    Energy Technology Data Exchange (ETDEWEB)

    Mikuš, Ján M., E-mail: JanMikus.nrc@hotmail.com

    2014-02-15

    Highlights: • Power (fission rate) distribution gradients can represent fuel failure root causes. • Positions with above gradients were investigated in WWER type cores on reactor LR-0. • Above gradients were evaluated near core heterogeneities and construction materials. • Results can be used for code validation and fuel failure occurrence investigation. - Abstract: Neutron flux non-uniformity and gradients of neutron current resulting in corresponding power (fission rate) distribution changes can represent root causes of the fuel failure. Such situation can be expected in vicinity of some core heterogeneities and construction materials. Since needed data cannot be obtained from nuclear power plant (NPP), results of some benchmark type experiments performed on light water, zero-power research reactor LR-0 were used for investigation of the above phenomenon. Attention was focused on determination of the spatial power distribution changes in fuel assemblies (FAs): Containing fuel rods (FRs) with Gd burnable absorber in WWER-440 and WWER-1000 type cores, Neighboring the core blanket and dummy steel assembly simulators on the periphery of the WWER-440 standard and low leakage type cores, resp., Neighboring baffle in WWER-1000 type cores, and Neighboring control rod (CR) in WWER-440 type cores, namely (a) power peak in axial power distribution in periphery FRs of the adjacent FAs near the area between CR fuel part and butt joint to the CR absorbing part and (b) decrease in radial power distribution in FRs near CR absorbing part. An overview of relevant experimental results from reactor LR-0 and some information concerning leaking FAs on NPP Temelín are presented. Obtained data can be used for code validation and subsequently for the fuel failure occurrence investigation.

  4. Measurements and observations on microscopic swelling in MX-type fuels

    International Nuclear Information System (INIS)

    Ronchi, C.; Ray, I.L.F.; Thiele, H.; Laar, J. van de.

    1978-01-01

    Microscopic swelling has been investigated by electron microscopy in several MX-type fuels, irradiated in fast and thermal neutron flux. The results show that fission gas bubbles in these compounds grow to large sizes if the in-pile fuel temperature rises above a critical value (swelling critical temperature Tsub(C)). A comparison has been made of the swelling rates in fuels of different composition, showing that Tsub(C) increases from carbides to nitrides. In fuels subjected to in-pile restructuring (highly rated) He-bonded pins microscopic swelling is affected by pore and grain boundary migration. The influence of these phenomena on the fuel swelling performance has been discussed

  5. Re-qualification of MTR-type fuel plates fabrication process

    International Nuclear Information System (INIS)

    Elseaidy, I.M.; Ghoneim, M.M.

    2010-01-01

    The fabricability issues with increased uranium loading due to use low enrichment of uranium (LEU), i.e. less than 20 % of U 235 , increase the problems which occur during compact manufacturing, roll bonding of the fuel plates, potential difficulty in forming during rolling process, mechanical integrity of the core during fabrication, potential difficulty in meat homogeneity, and the ability to fabricate plates with thicker core as a means of increasing total uranium loading. To produce MTR- type fuel plates with high uranium loading (HUL) and keep the required quality of these plates, many of qualification process must be done in the commissioning step of fuel fabrication plant. After that any changing of the fabrication parameters, for example changing of any of the raw materials, devises, operators, and etc., a re- qualification process should be done in order to keep the quality of produced plates. Objective of the present work is the general description of the activities to be accomplished for re-qualification of manufacturing MTR- type nuclear fuel plates. For each process to be re-qualified, a detailed of re-qualification process were established. (author)

  6. A Favela Santa Marta e seus guias de turismo: identidade, mobilização e conflito

    Directory of Open Access Journals (Sweden)

    Gabriel Ferreira Barbosa

    2015-03-01

    Full Text Available A proposta deste artigo é descrever o processo de constituição de um mercado turístico na Favela Santa Marta e os efeitos observados no que se refere aos conflitos e consequentes mobilizações que reivindicam a legitimidade de se contar a história desse lugar. Por ter sido a primeira favela da cidade do Rio de Janeiro onde foi instalada uma Unidade de Polícia Pacificadora (UPP em dezembro de 2008[1], a Santa Marta passa a ser reconhecida como uma “favela modelo” por parte da administração pública, como se ela fosse um “laboratório” para experiências de políticas públicas. Esse processo ocorre em um contexto – notadamente os megaeventos esportivos como a Copa do Mundo FIFA 2014 e as Olimpíadas de 2016 -  em que se busca um “reencantamento” das representações sobre as favelas, mais especificamente aquelas onde as UPPs passam a fazer parte do cotidiano de seus moradores. O programa Rio Top Tour, que é a política analisada aqui, foi lançado na Favela Santa Marta, e consistiu na formação de guias turísticos locais e estímulo ao desenvolvimento de um mercado turístico. A partir da atuação desses guias e a sua mobilização para serem os legítimos portadores da narrativa sobre a história da Favela Santa Marta no âmbito do mercado turístico local, analisarei os conflitos que resultam desse processo. Antes disso, entretanto, é preciso apresentar este lugar e as mudanças que vêm ocorrendo nos últimos anos desde a instalação da UPP – que é o corte temporal proposto neste trabalho.

  7. CFD thermal-hydraulic analysis of a CANDU fuel channel with SEU43 type fuel bundle

    International Nuclear Information System (INIS)

    Catana, A.; Prisecaru, Ilie; Dupleac, D.; Danila, Nicolae

    2009-01-01

    This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational Fluid Dynamics) methodology approach, when SEU43 fuel bundles are used. Comparisons with STD37 fuel bundles are done in order to evaluate the influence of geometrical differences of the fuel bundle types on fluid flow properties. We adopted a strategy to analyze only the significant segments of fuel channel, namely : - the fuel bundle junctions with adjacent segments; - the fuel bundle spacer planes with adjacent segments; - the fuel bundle segments with turbulence enhancement buttons; - and the regular segments of fuel bundles. The computer code used is an academic version of FLUENT code, available from UPB. The complex flow domain of fuel bundles contained in pressure tube and operating conditions determine a high turbulence flow and in some parts of fuel channel also a multi-phase flow. Numerical simulation of the flow in the fuel channel has been achieved by solving the equations for conservation of mass, momentum and energy. For turbulence model the standard k-model is employed although other turbulence models can be used. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. of a SEU43 fuel bundle in conditions of a typical CANDU 6 fuel channel starting from some experience gained in a previous work. (authors)

  8. Model development of UO_2-Zr dispersion plate-type fuel behavior at early phase of severe accident and molten fuel meat relocation

    International Nuclear Information System (INIS)

    Zhang Zhuohua; Yu Junchong; Peng Shinian

    2014-01-01

    According to former study on oxygen diffusion, Nb-Zr solid reaction and UO_2-Zr solid reaction, the models of oxidation, solid reaction in fuel meat and relocation of molten fuel meat are developed based on structure and material properties of UO_2-Zr dispersion plate-type fuel, The new models can supply theoretical elements for the safety analysis of the core assembled with dispersion plate-type fuel under severe accident. (authors)

  9. Refining fuels of the heavy gas--oil type

    Energy Technology Data Exchange (ETDEWEB)

    Bruzac, J F.A.

    1930-01-28

    This invention has for its object the production of a new type of gas-oil fuel, obtained from crude petroleum, shale oil, and peat oil, according to the method of treatment mentioned, by means of which is obtained from gas oil, shale oil, lignite oil, and peat oil (deprived of asphaltic, and bituminous, resinous, and sulfur compounds), a fuel suitable for running Diesel, Junkers, and Clerget motors and all others of the same kind, by diminishing considerably the fouling and attack on the metal.

  10. Fuel type characterization and potential fire behavior estimation in Sardinia and Corsica islands

    Science.gov (United States)

    Bacciu, V.; Pellizzaro, G.; Santoni, P.; Arca, B.; Ventura, A.; Salis, M.; Barboni, T.; Leroy, V.; Cancellieri, D.; Leoni, E.; Ferrat, L.; Perez, Y.; Duce, P.; Spano, D.

    2012-04-01

    Wildland fires represent a serious threat to forests and wooded areas of the Mediterranean Basin. As recorded by the European Commission (2009), during the last decade Southern Countries have experienced an annual average of about 50,000 forest fires and about 470,000 burned hectares. The factor that can be directly manipulated in order to minimize fire intensity and reduce other fire impacts, such as three mortality, smoke emission, and soil erosion, is wildland fuel. Fuel characteristics, such as vegetation cover, type, humidity status, and biomass and necromass loading are critical variables in affecting wildland fire occurrence, contributing to the spread, intensity, and severity of fires. Therefore, the availability of accurate fuel data at different spatial and temporal scales is needed for fire management applications, including fire behavior and danger prediction, fire fighting, fire effects simulation, and ecosystem simulation modeling. In this context, the main aims of our work are to describe the vegetation parameters involved in combustion processes and develop fire behavior fuel maps. The overall work plan is based firstly on the identification and description of the different fuel types mainly affected by fire occurrence in Sardinia (Italy) and Corsica (France) Islands, and secondly on the clusterization of the selected fuel types in relation to their potential fire behavior. In the first part of the work, the available time series of fire event perimeters and the land use map data were analyzed with the purpose of identifying the main land use types affected by fires. Thus, field sampling sites were randomly identified on the selected vegetation types and several fuel variables were collected (live and dead fuel load partitioned following Deeming et al., (1977), depth of fuel layer, plant cover, surface area-to-volume ratio, heat content). In the second part of the work, the potential fire behavior for every experimental site was simulated using

  11. Feasibility of Electromagnetic Acoustic Evaluation for Quality Test of a Plate-type Nuclear Fuel

    International Nuclear Information System (INIS)

    Jung, Hyun Kyu; Lee, Yoon Sang; Cheong, Yong Moo

    2010-01-01

    Most research and test reactors use the nuclear fuel plates which are consisted of a fuel core in aluminum alloy. Recently KAERI signed a deal with the Jordan Atomic Energy Commission to build the research reactor and have to supply the plate-type nuclear fuels. For the demands of world market, KAERI started the research and development of the plate-type fuel elements and endeavored to achieve a localization of fuel fabrication. For the inspection of plate-type fuel elements to be used in Research Reactors, an immersion pulse-echo ultrasonic technique was applied. This inspection was done with water, so a nuclear fuel was immersed to be prone to corrosion and needed to have time and cost due to an additional process. The sample that will be examined within this paper is a non-ferromagnetic material such as aluminum which has a good acousto-elastic property, for an effective inspection of a bond quality for a nuclear fuel under a manufacturing environment. The purpose of this study is to investigate the feasibility of an EMAT technology for an automated inspection of a nuclear fuel without water

  12. Operation of Nuclear Fuel Based on Reprocessed Uranium for VVER-type Reactors in Competitive Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Troyanov, V.; Molchanov, V.; Tuzov, A. [TVEL Corporation, 49 Kashirskoe shosse, Moscow 115409 (Russian Federation); Semchenkov, Yu.; Lizorkin, M. [RRC ' Kurchatov Institute' (Russian Federation); Vasilchenko, I.; Lushin, V. [OKB ' Gidropress' (Russian Federation)

    2009-06-15

    Current nuclear fuel cycle of Russian nuclear power involves reprocessed low-enriched uranium in nuclear fuel production for some NPP units with VVER-type LWR. This paper discusses design and performance characteristics of commercial nuclear fuel based on natural and reprocessed uranium. It presents the review of results of commercial operation of nuclear fuel based on reprocessed uranium on Russian NPPs-unit No.2 of Kola NPP and unit No.2 of Kalinin NPP. The results of calculation and experimental validation of safe fuel operation including necessary isotope composition conformed to regulation requirements and results of pilot fuel operation are also considered. Meeting the customer requirements the possibility of high burn-up achieving was demonstrated. In addition the paper compares the characteristics of nuclear fuel cycles with maximum length based on reprocessed and natural uranium considering relevant 5% enrichment limitation and necessity of {sup 236}U compensation. The expedience of uranium-235 enrichment increasing over 5% is discussed with the aim to implement longer fuel cycles. (authors)

  13. Modeling solid-fuel dispersal during slow loss-of-flow-type transients

    International Nuclear Information System (INIS)

    DiMelfi, R.J.; Fenske, G.R.

    1981-01-01

    The dispersal, under certain accident conditions, of solid particles of fast-reactor fuel is examined in this paper. In particular, we explore the possibility that solid-fuel fragmentation and dispersal can be driven by expanding fission gas, during a slow LOF-type accident. The consequences of fragmentation are studied in terms of the size and speed of dispersed particles, and the overall quantity of fuel moved. (orig.)

  14. Evaluation of plate type fuel elements by eddy current test method

    International Nuclear Information System (INIS)

    Frade, Rangel Teixeira

    2015-01-01

    Plate type fuel elements are used in MTR research nuclear reactors. The fuel plates are manufactured by assembling a briquette containing the fissile material inserted in a frame, with metal plates in both sides of the set, to act as a cladding. This set is rolled under controlled conditions in order to obtain the fuel plate. In Brazil, this type of fuel is manufactured by IPEN and used in the IEA-R1 reactor. After fabrication of three batches of fuel plates, 24 plates, one of them is taken, in order to verify the thickness of the cladding. For this purpose, the plate is sectioned and the thickness measurements are carried out by using optical microscopy. This procedure implies in damage of the plate, with the consequent cost. Besides, the process of sample preparation for optical microscopy analysis is time consuming, it is necessary an infrastructure for handling radioactive materials and there is a generation of radioactive residues during the process. The objective of this study was verify the applicability of eddy current test method for nondestructive measurement of cladding thickness in plate type nuclear fuels, enabling the inspection of all manufactured fuel plates. For this purpose, reference standards, representative of the cladding of the fuel plates, were manufactured using thermomechanical processing conditions similar to those used for plates manufacturing. Due to no availability of fuel plates for performing the experiments, the presence of the plate’s core was simulated using materials with different electrical conductivities, fixed to the thickness reference standards. Probes of eddy current testing were designed and manufactured. They showed high sensitivity to thickness variations, being able to separate small thickness changes. The sensitivity was higher in tests performed on the reference standards and samples without the presence of the materials simulating the core. For examination of the cladding with influence of materials simulating the

  15. A dynamic, dependent type system for nuclear fuel cycle code generation

    Energy Technology Data Exchange (ETDEWEB)

    Scopatz, A. [The University of Chicago 5754 S. Ellis Ave, Chicago, IL 60637 (United States)

    2013-07-01

    The nuclear fuel cycle may be interpreted as a network or graph, thus allowing methods from formal graph theory to be used. Nodes are often idealized as nuclear fuel cycle facilities (reactors, enrichment cascades, deep geologic repositories). With the advent of modern object-oriented programming languages - and fuel cycle simulators implemented in these languages - it is natural to define a class hierarchy of facility types. Bright is a quasi-static simulator, meaning that the number of material passes through a facility is tracked rather than natural time. Bright is implemented as a C++ library that models many canonical components such as reactors, storage facilities, and more. Cyclus is a discrete time simulator, meaning that natural time is tracked through out the simulation. Therefore a robust, dependent type system was developed to enable inter-operability between Bright and Cyclus. This system is capable of representing any fuel cycle facility. Types declared in this system can then be used to automatically generate code which binds a facility implementation to a simulator front end. Facility model wrappers may be used either internally to a fuel cycle simulator or as a mechanism for inter-operating multiple simulators. While such a tool has many potential use cases it has two main purposes: enabling easy performance of code-to-code comparisons and the verification and the validation of user input.

  16. Sistemática para avaliação da conformidade: uma proposta para uso de cão-guia

    OpenAIRE

    Brito, Joana D'Arc de

    2009-01-01

    O objetivo do presente estudo é identificar e propor quais os principais requisitos para a Avaliação dos Centros de Treinamento, de Treinadores e dos Instrutores Autônomos de cães-guia, de forma a atender o que prescreve a Lei n.º 11.126, de 27 de junho de 2005 no concernente às necessidades de adaptação dos segmentos envolvidos. Esta pesquisa foi estruturada em 3 (três) etapas. No primeiro momento utilizou-se de fontes bibliográficas para o levantamento de obras de referência para o desenvol...

  17. J.S.Bach variações "Goldberg" : um guia para a formação do homem completo

    OpenAIRE

    Helena Jank

    1988-01-01

    Resumo: Dentre muitas formas possíveis de abordagem para esta obra monumental de J. S. Bach, o presente trabalho traz um enfoque eminentemente humanístico, através do qual transparece a intenção do compositor de estruturar, na forma de tema e variações um ?guia para a formação do homem completo? Esta intenção se manifesta na organização geral da obra, com uma clara divisão em três grupos de variações, nos quais são explorados aspectos técnicos, composicionais e de interpretação. Com base na ...

  18. Fuel management of mixed reactor type power plant systems

    International Nuclear Information System (INIS)

    Csom, Gyula

    1988-01-01

    In equilibrium symbiotic power plant system containing both thermal reactors and fast breeders, excess plutonium produced by the fast breeders is used to enrich the fuel of the thermal reactors. In plutonium deficient symbiotic power plant system plutonium is supplied both by thermal plants and fast breeders. Mathematical models were constructed and different equations solved to characterize the fuel utilization of both systems if they contain only a single thermal type and a single fast type reactor. The more plutonium is produced in the system, the higher output ratio of thermal to fast reactors is achieved in equilibrium symbiotic power plant system. Mathematical equations were derived to calculate the doubling time and the breeding gain of the equilibrium symbiotic system. (V.N.) 2 figs.; 2 tabs

  19. Perspectives for practical application of the combined fuel kernels in VVER-type reactors

    International Nuclear Information System (INIS)

    Baranov, V.; Ternovykh, M.; Tikhomirov, G.; Khlunov, A.; Tenishev, A.; Kurina, I.

    2011-01-01

    The paper considers the main physical processes that take place in fuel kernels under real operation conditions of VVER-type reactors. Main attention is given to the effects induced by combinations of layers with different physical properties inside of fuel kernels on these physical processes. Basic neutron-physical characteristics were calculated for some combined fuel kernels in fuel rods of VVER-type reactors. There are many goals in development of the combined fuel kernels, and these goals define selecting the combinations and compositions of radial layers inside of the kernels. For example, the slower formation of the rim-layer on outer surface of the kernels made of enriched uranium dioxide can be achieved by introduction of inner layer made of natural or depleted uranium dioxide. Other potential goals (lower temperature in the kernel center, better conditions for burn-up of neutron poisons, better retention of toxic materials) could be reached by other combinations of fuel compositions in central and peripheral zones of the fuel kernels. Also, the paper presents the results obtained in experimental manufacturing of the combined fuel pellets. (authors)

  20. Field experience of new nuclear fuel types on the Kola NPP

    International Nuclear Information System (INIS)

    Adeev, V.; Burlov, S.; Panov, A.; Saprykin, V.

    2008-01-01

    Specificity of the Kola nuclear power plant geographical position, conditions of region economics determine fuel management strategy. Isolation of Kola power supply system and, as a consequence, generating capacities redundancy cause operation of the nuclear power plant on reduced power level. At the same time there is a need to operate the power unit on the maximum power level in the case of not planned conditions. The basis of in-core fuel management is an achievement of the maximal burnup under providing of high installed capacity. At present there are not abilities to improve the fuel cycle based on traditional implementation fuel assemblies. Burnup maximum in these fuel cycles is achieved. At the core periphery installed highest possible quantity of the burned-up assemblies in the view of safety operation margins satisfaction. Works on application of the second generation fuel have been carried out on the Kola NPP since 2002. Fuel assemblies of this type are profiled. Burnable absorber, changed lattice spacing in relation to standard fuel, changed height of a fuel column, thickness of fuel pin clad are applied. In CR fuel followers modernized docking unit (with hafnium plates are intended for energy-release splash suppression) is used. At present 2-nd generation fuel is in experimental operation on unit 3 (18-21 fuel cycles, 2002-2007 years) and unit 4 (18-19 fuel cycles, 2005-2007 years). Safety margins did not exceeded. Coolant activity did not exceed the limiting value. There were not damaged fuel assemblies of second generation. Originally in the project of applications of new fuel it was supposed to refuel annually 78 fresh assemblies. At the moment annual refueling consists of 66 assemblies with effective enrichment 3.82 %. Cycle duration does not exceed 250-260 effective days. The part of assemblies is left on 5-th cycle of operation. In a similar fuel cycle in 2007 on the unit 1 operation with profiled fuel (enrichment of 3.82 %) of shakeproof type

  1. Guia para prática clínica: fisioterapia em pacientes com Doença Pulmonar Obstrutiva Crônica (DPOC Clinical practice guideline for physical therapy in patients with Chronic Obstructive Pulmonary Disease (COPD: portuguese version

    Directory of Open Access Journals (Sweden)

    D Langer

    2009-06-01

    Full Text Available INTRODUÇÃO: No contexto da colaboração internacional para desenvolvimento de guias práticos (ou guidelines, a Sociedade Real Holandesa de Fisioterapia (Koninklijk Nederlands Genootschap voor Fysiotherapie, KNGF se propôs a desenvolver um guia para esclarecimento sobre a prática clínica de Fisioterapia em pacientes com Doença Pulmonar Obstrutiva Crônica (DPOC, assim como também optou por estimular a sua tradução para outras línguas, a fim de torná-lo acessível para públicos internacionais. OBJETIVOS: O presente guia é a versão em língua portuguesa do Guia para Prática Clínica de Fisioterapia em pacientes com DPOC desenvolvido pela KNGF, que teve como objetivo descrever a Fisioterapia baseada em evidências para pacientes com DPOC que apresentam limitação da função pulmonar, da função muscular respiratória e periférica, da capacidade de exercício, da depuração mucociliar e da qualidade de vida, além de limitações em relação à atividade física na vida diária pela dispneia e/ou intolerância ao exercício. CONCLUSÃO: O guia propõe-se principalmente a prover recomendações terapêuticas práticas que auxiliem o fisioterapeuta a oferecer o melhor tratamento possível para pacientes com DPOC, consideradas as evidências científicas disponíveis na atualidade.INTRODUCTION: In the context of international collaboration for the development of practice guidelines, the Royal Dutch Society for Physical Therapy (Koninklijk Nederlands Genootschap voor Fysiotherapie, KNGF has developed guidelines for the clinical practice of physical therapy in patients with Chronic Obstructive Pulmonary Disease (COPD. It has also stimulated its translation into other languages to make it accessible to international audiences. OBJECTIVES: The present document brings the Portuguese version of the KNGF Clinical Practice Guidelines for physical therapy in COPD patients. Its purpose was to describe evidence-based physical therapy for

  2. Study on new-type fuel-related assembly handling tools for PWR NPP

    International Nuclear Information System (INIS)

    Fan Xiumei

    2013-01-01

    This article describes the design and study on a set of new-type fuel-related assembly snatching tools used for PWR NPP. The purpose is mainly to enhance the tool safety, reliability and convenientness by improvement of the mechanism and structure of the tool for snatching preciseness and avoiding from falling and abrasion of fuel-related assemblies for any condition. The new-type fuel-related assembly handling tools are compared with similar equipment in worldwide in terms of function, main technical characteristic, and safety and protection, some of them are better than the similar equipment in that they have reliable loading and unloading and conveying capabilities. (author)

  3. Study of behavior of cermet fuel elements on IGR reactor under RIA type accident condition

    International Nuclear Information System (INIS)

    Vasil'ev, Yu.S.; Vurim, A.D.; Koltyshev, S.M.; Pakhnits, V.A.; Tukhvatulin, Sh.T.; Popov, V.V.; Ryzhkov, A.N.

    1996-01-01

    In 1993 December in IGR reactor of Inst. of Atomic Energy of National Nuclear Center of Republic of Kazakstan the second batch of in-pile testing of perspective cermet fuel elements under the condition, simulating RIA type accident was conducted. In the second batch of testing during eight start-ups 10 cermet fuel elements were examined. Among which 8 of monolith type and 2 fuel elements with false jacket beside cladding (FJF), as well as, 6 standard fuel elements of WWER-1000 type reactor with dioxide fuel were tested. 2 fuel elements - cermet and standard were placed into capsule filled with water. To measure energy release for the each start-up two fission monitor and inside core control gauge were placed. In all the start-ups operation mode of IGR was neutron pulse. Power of fuel element kept changing from 151 to 336 k W; energy release was 38-93 kJ/gr m 235 U; maximum temperature of cermet fuel was 1943-2173 K, of dioxide fuel - 1923-2843 K. The testing has demonstrated that operability of cermet fuel elements under reactivity accident condition with pulse width of 0,2 s is, at least, not less that operability of dioxide fuel elements, through advantages of cermet fuel under these conditions are revealed to the least extent

  4. Core conversion from rod to plate type fuel elements in research reactors

    International Nuclear Information System (INIS)

    Khattab, M.S.; Mina, A.R.

    1997-01-01

    Core thermalhydraulic analysis have been performed for rod and plate types fuel elements without altering the core bundles square grid spacer (68 mm, side) and coolant mass flow rate. The U O 2 -Mg, 10% enrichment rod type fuel elements are replaced by the MTR plate type, U-Al alloy of 20% enrichment. Coolant mass flux increased from 2000 kg/m 2 S to 5000 kg/m 2 S. Reactor power could be upgraded from 2 to 10 MW without significantly altering the steady state, thermal-hydraulic safety margins. Fuel, clad and coolant transient temperatures are determined inside the core hot channel during flow coast down using paret code. Residual heat removal system of 20% coolant capacity is necessary for upgrading reactor power to encounter the case of pumps off at 10 MW nominal operation. 6 figs., 2 tabs

  5. Performance evaluation of the Loviisa advanced type fuel rods

    International Nuclear Information System (INIS)

    Ranta-Puska, K.; Pihlatie, M.

    2001-01-01

    The fuel vendor TVEL has supplied to Loviisa WWER-440 power plant six lead assemblies of an advanced type which have profiling of the fuel enrichment, demountability of the assembly and a reduced shroud wall thickness. The pool side examination programme of these assemblies is underway including visual inspections, diameter and length measurements between operation cycles, and end-of-life fission gas release measurements, determined from 85 Kr activity in the plenum. Complementary evaluations and testing of models are done with the ENIGMA fuel performance code. The diameters of the corner rods have decreased to 30 μm during the first cycle and 40 to 70 μm after two cycles (with rod burnups of 24-30 MWd/kgU). The extent of creep-down is generally as expected, and agrees with the creep model adjusted for Russian Zr1%Nb cladding type and the Loviisa coolant and neutron flux conditions. The gap closure and reversed hoop strain are to be awaited during the third cycle so the new data will be an interesting validation exercise for the model and ENIGMA. Calculated temperatures stay low, and therefore low fission gas release fractions are anticipated as well

  6. Crossflow characteristics of flange type fuel element for very high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Takizuka, Takakazu; Kaburaki, Hideo; Suzuki, Kunihiko; Nakamura, Masahide.

    1987-01-01

    Fuel element design incorporating mating flanges at block end faces has the potential to improve thermal hydraulic performance of a VHTR (very high temperature gas-cooled reactor) core. As part of research and development efforts to establish flange type fuel element design, experiments and analyses were carried out on crossflow through interface gap between elements. Air at atmospheric pressure and ambient temperature was used as a fluid. Crossflow loss coefficient factors were obtained with three test models, having different flange mating clearances, for various interface gap configurations, gap widths and block misalignments. It was found that crossflow loss coefficient factors for flange type fuel element were much larger than those for conventional flat-faced element. Numerical analyses were also made using a simple model devised to represent the crossflow path at the fuel element interface. The close agreement between numerical results and experimental data indicated that this model could predict well the crossflow characteristics of the flange type fuel element. (author)

  7. Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

    Science.gov (United States)

    Permana, Sidik

    2017-07-01

    A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

  8. Calculation device for fuel power history in BWR type reactors

    International Nuclear Information System (INIS)

    Sakagami, Masaharu.

    1980-01-01

    Purpose: To enable calculations for power history and various variants of power change in the power history of fuels in a BWR type reactor or the like. Constitution: The outputs of the process computation for the nuclear reactor by a process computer are stored and the reactor core power distribution is judged from the calculated values for the reactor core power distribution based on the stored data. Data such as for thermal power, core flow rate, control rod position and power distribution are recorded where the changes in the power distribution exceed a predetermined amount, and data such as for thermal power and core flow rate are recorded where the changes are within the level of the predetermined amount, as effective data excluding unnecessary data. Accordingly, the recorded data are taken out as required and the fuel power history and the various variants in the fuel power are calculated and determined in a calculation device for fuel power history and variants for fuel power fluctuation. (Furukawa, Y.)

  9. Post-pulse detail metallographic examinations of low-enriched uranium silicide plate-type miniature fuel

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1991-10-01

    Pulse irradiation at Nuclear Safety Research Reactor (NSRR) was performed using low-enriched (19.89 w% 235 U) unirradiated silicide plate-type miniature fuel which had a density of 4.8 gU/cm 3 . Experimental aims are to understand the dimensional stability and to clarify the failure threshold of the silicide plate-type miniature fuel under power transient conditions through post-pulse detail metallographic examinations. A silicide plate-type miniature fuel was loaded into an irradiation capsule and irradiated by a single pulse. Deposited energies given in the experiments were 62, 77, 116 and 154 cal/g·fuel, which lead to corresponding peak fuel plate temperatures, 201 ± 28degC, 187 ± 10degC, 418 ± 74degC and 871 ± 74degC, respectively. Below 400degC, reliability and dimensional stability of the silicide plate fuel was sustained, and the silicide plate fuel was intact. Up to 540degC, wall-through intergranular crackings occurred in the Al-3%Mg alloy cladding. With the increase of the temperature, the melting of the aluminum cladding followed by recrystallization, the denudation of fuel core and the plate-through intergranular cracking were observed. With the increase of the temperature beyond 400degC, the bowing of fuel plate became significant. Above the temperature of 640degC molten aluminum partially reacted with the fuel core, partially flowed downward under the influence of surface tension and gravity, and partially formed agglomerations. Judging from these experimental observations, the fuel-plate above 400degC tends to reduce its dimensional stability. Despite of the apparent silicide fuel-plate failure, neither generation of pressure pulse nor that of mechanical energy occurred at all. (J.P.N.)

  10. Performance evaluation of CPF shredder type mechanical crusher with simulated core fuel pin

    International Nuclear Information System (INIS)

    Nakahara, Masaumi; Sano, Yuichi; Aose, Shin-ichi

    2006-12-01

    In the advanced aqueous reprocessing system, powder fuel dissolution has been investigated, which is quite effective on the dissolution for highly concentrated solution. As one of the effective means that powder the irradiated MOX fuel, we have been developing shredder type mechanical crusher. This apparatus can automatically crush the sheared fuel pieces by twin-shaft disk blades, powder the crushed fragments by disk blades and screen blade, and recover the powdered fuel. The shredder type mechanical crusher was developed for using in a hot cell in Chemical Processing Facility, and the first crush experiment with this crusher was carried out at July 2004 using the simulated core fuel pin. This experiment showed that the crushed fragments could not be grinded efficiency because screen blade vibrated up and down during the operation. Additionally, the strength of screen blade block was insufficient to crush the sheared fuel pieces stably. Therefore, about 70% of fuel was recovered in maximum. Based on the results of the first experiment, screen blade was fixed up mainly and the second experiment was carried out with improved apparatus at September 2005. In this experiment, about 96% of fuel could be recovered in maximum because screen blade was stable during the operation. (J.P.N.)

  11. Bi-fuel System - Gasoline/LPG in A Used 4-Stroke Motorcycle - Fuel Injection Type

    Science.gov (United States)

    Suthisripok, Tongchit; Phusakol, Nachaphat; Sawetkittirut, Nuttapol

    2017-10-01

    Bi-fuel-Gasoline/LPG system has been effectively and efficiently used in gasoline vehicles with less pollutants emission. The motorcycle tested was a used Honda AirBlade i110 - fuel injection type. A 3-litre LPG storage tank, an electronic fuel control unit, a 1-mm LPG injector and a regulator were securely installed. The converted motorcycle can be started with either gasoline or LPG. The safety relief valve was set below 48 kPa and over 110 kPa. The motorcycle was tuned at the relative rich air-fuel ratio (λ) of 0.85-0.90 to attain the best power output. From dynamometer tests over the speed range of 65-100 km/h, the average power output when fuelling LPG was 5.16 hp; dropped 3.9% from the use of gasoline91. The average LPG consumption rate from the city road test at the average speed of 60 km/h was 40.1 km/l, about 17.7% more. This corresponded to lower LPG’s energy density of about 16.2%. In emission, the CO and HC concentrations were 44.4% and 26.5% lower. Once a standard gas equipment set with ECU and LPG injector were securely installed and the engine was properly tuned up to suit LPG’s characteristics, the converted bi-fuel motorcycle offers efficiently, safely and economically performance with environmental friendly emission.

  12. Development of simulation code for FBR spent fuel dissolution with rotary drum type continuous dissolver

    International Nuclear Information System (INIS)

    Sano, Yuichi; Katsurai, Kiyomichi; Washiya, Tadahiro; Koizumi, Tsutomu; Matsumoto, Satoshi

    2011-01-01

    Japan Atomic Energy Agency (JAEA) has been studying rotary drum type continuous dissolver for FBR spent fuel dissolution. For estimating the fuel dissolution behavior under several operational conditions in this dissolver, we have been developing the simulation code, PLUM, which mainly consists of 3 modules for calculating chemical reaction, mass transfer and thermal balance in the rotary drum type continuous dissolver. Under the various conditions where dissolution experiments were carried out with the batch-wise dissolver for FBR spent fuel and with the rotary drum type continuous dissolver for UO 2 fuel, it was confirmed that the fuel dissolution behaviors calculated by the PLUM code showed good agreement with the experimental ones. Based on this result, the condition for obtaining the dissolver solution with high HM (heavy metal : U and Pu) concentration (∼500g/L), which is required for the next step, i.e. crystallization process, was also analyzed by this code and appropriate operational conditions with the rotary drum type continuous dissolver, such as feedrate, concentration and temperature of nitric acid, could be clarified. (author)

  13. Enrichment measurement in TRIGA type fuels; Medicion de enriquecimiento en combustibles tipo Triga

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-05-15

    The Department of Energy of the United States of North America, through the program 'Idaho Operations Nuclear Spent Fuel Program' of the Idaho National Engineering and Environmental Laboratory (INEEL), in Idaho Falls; Idaho USA, hires to Global Technologies Inc. (GTI) to develop a prototype device of detection enrichment uranium (DEU Detection of Enrichment of Uranium) to determine quantitatively the enrichment in remainder U-235 in a TRIGA fuel element at the end of it useful life. The characteristics of the prototype developed by GTI are the following ones: It allows to carry out no-destructive measurements of TRIGA type fuel. Easily transportable due to that reduced of it size. The determination of the enrichment (in grams of U-235) it is obtained with a precision of 5%. The National Institute of Nuclear Research (ININ), in its facilities of the Nuclear Center of Mexico, it has TRIGA type fuel of high and low enrichment (standard and FLIP) fresh and with burnt, it also has the infrastructure (hot cells, armor-plating of transport, etc) and qualified personnel to carry out the necessary maneuvers to prove the operation of the DEU prototype. For this its would be used standard type fuel elements and FLIP, so much fresh as with certain burnt one. In the case of the fresh fuels the measurement doesn't represent any risk, the fuels before and after the measurement its don't contain a quantity of fission products that its represent a radiological risk in its manipulation; but in the case of the fuels with burnt the handling of the same ones represents an important radiological risk reason why for its manipulation it was used the transport armor-plating and the hot cells. (Author)

  14. Is fuel poverty in Ireland a distinct type of deprivation?

    OpenAIRE

    Watson, Dorothy; Maitre, Bertrand

    2014-01-01

    In this paper, we draw on the Central Statistics Office SILC data for Ireland to ask whether fuel poverty is a distinctive type of deprivation that warrants a fundamentally different policy response than poverty in general. We examine the overlap between fuel poverty (based on three self-report items) and poverty in general – with a particular emphasis on the national indicator of basic deprivation which is used in the measurement of poverty for policy purposes in Ireland. We examine changes ...

  15. Status of fuel element technology for plate type dispersion fuels with high uranium density

    International Nuclear Information System (INIS)

    Hrovat, M.; Huschka, H.; Koch, K.H.; Nazare, S.; Ondracek, G.

    1983-01-01

    A number of about 20 Material Test and Research Reactors in Germany and abroad is supplied with fuel elements by the company NUKEM. The power of these reactors differs widely ranging from up to about 100 MW. Consequently, the uranium density of the fuel elements in the meat varies considerably depending on the reactor type and is usually within the range from 0.4 to 1.3 g U/cm 3 if HEU is used. In order to convert these reactors to lower uranium enrichment (19.75% 235-U) extensive work is carried out at NUKEM since about two years with the goal to develop fuel elements with high U-density. This work is sponsored by the German Ministry for Research and Technology in the frame of the AF-program. This paper reports on the present state of development for fuel elements with high U-density fuels at NUKEM is reported. The development works were so far concentrated on UAl x , U 3 O 8 and UO 2 fuels which will be described in more detail. In addition fuel plates with new fuels like e.g. U-Si or U-Fe compounds are developed in collaboration with KfK. The required uranium densities for some typical reactors with low, medium, and high power are listed allowing a comparison of HEU and LEU uranium density requirements. The 235-U-content in the case of LEU is raised by 18%. Two different meat thicknesses are considered: Standard thickness of 0.5 mm; and increased thickness of 0.76 mm. From this data compilation the objective follows: in the case of conversion to LEU (19.75% 235-U-enrichment), uranium densities have to be made available up to 24 gU/cm 3 meat for low power level reactors, up to 33 gU/cm 3 meat for medium power level reactors, and between 5.75 and 7.03 g/cm 3 meat for high power level reactors according to this consideration

  16. Fast reactor parameter optimization taking into account changes in fuel charge type during reactor operation time

    International Nuclear Information System (INIS)

    Afrin, B.A.; Rechnov, A.V.; Usynin, G.B.

    1987-01-01

    The formulation and solution of optimization problem for parameters determining the layout of the central part of sodium cooled power reactor taking into account possible changes in fuel charge type during reactor operation time are performed. The losses under change of fuel composition type for two reactor modifications providing for minimum doubling time for oxide and carbide fuels respectively, are estimated

  17. Fuel assembly for pressure loss variable PWR type reactor

    International Nuclear Information System (INIS)

    Yoshikuni, Masaaki.

    1993-01-01

    In a PWR type reactor, a pressure loss control plate is attached detachably to a securing screw holes on the lower surface of a lower nozzle to reduce a water channel cross section and increase a pressure loss. If a fuel assembly attached with the pressure loss control plate is disposed at a periphery of the reactor core where the power is low and heat removal causes no significant problem, a flowrate at the periphery of the reactor core is reduced. Since this flowrate is utilized for removal of heat from fuel assemblies of high powder at the center of the reactor core where a pressure loss control plate is not attached, a thermal limit margin of the whole reactor core is increased. Thus, a limit of power peaking can be moderated, to obtain a fuel loading pattern improved with neutron economy. (N.H.)

  18. Impact of Fuel Type on the Internal Combustion Engine Condition

    Directory of Open Access Journals (Sweden)

    Zdravko Schauperl

    2012-07-01

    Full Text Available The paper studies the influence of liquefied petroleum gas as alternative fuel on the condition of the internal combustion engine. The traffic, energy, economic and ecological influence as well as the types of fuel are studied and analyzed in an unbiased manner, objectively, and in detail, and the obtained results are compared with the condition of the engine of a vehicle powered by the stipulated fuel, petrol Eurosuper 95. The study was carried out on two identical passenger cars with one being fitted with gas installation. The obtained results show that properly installed gas installations in vehicles and the usage of LPG have no significant influence on the driving performances, but they affect significantly the ecological and economic parameters of using passenger cars.

  19. Analytical Evaluation to Determine Selected PAHs by HPLC in a Type 2 Fuel

    International Nuclear Information System (INIS)

    Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Escolano Segovia, O.; Garcia Frutos, F. J.

    2009-01-01

    An evaluation of analytical parameters to determine selected PAHs in a fuel oil type II by HPLC coupled to fluorescence and diode detectors is presented. The study was focused on four conventional treatments of these kinds of oil samples and the main objective was giving a measure of confidence level of PAH results in the fuel oil. This study was performed in the frame of the project Assessment of natural attenuation of PAHs in agricultural soil contaminated with fuel from an accidental spill (Spanish National Plain I+D+I, CTM2007-64537). This paper is presented as follows: Analysis of reference material 1582 (NIST) by using the four kinds of sample treatments of interest. Application of variance analysis to compare results obtained from type II fuel by using each sample treatment and chromatographic detector. Finally, a statistic calculation was performed to measure uncertainty components in chromatographic analysis. (Author)

  20. FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

    Science.gov (United States)

    Loeb, E.; Nicklas, J.H.

    1959-02-01

    A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

  1. Production of 15N for nitride type nuclear fuel

    International Nuclear Information System (INIS)

    Axente, Damian

    2005-01-01

    Full text: Nitride nuclear fuel is the choice for advanced nuclear reactors and ADS, considering its favorable properties as: melting point, excellent thermal conductivity, high fissile density, lower fission gas release and good radiation tolerance. The application of nitride fuels in different nuclear reactors requires use of 15 N enriched nitrogen to suppress 14 C production due to (n,p) reaction on 14 N. Nitride fuel is a promising candidate for transmutation in ADSs of radioactive minor actinides, which are converted into nitrides with 15 N for that purpose. Taking into account that at present the world wide 15 N market is about 20 - 40 Kg 15 N/y, the supply of that isotope for nitride type nuclear fuel, would demand an increase in production capacity by a factor of 1000. For an industrial plant producing 100 t/y 15 N at 99 at. % 15 N concentration, using present technology of 15 N/ 14 N isotopic exchange in Nitrox system, the first separation stage of the cascade would be fed with 10M HNO 3 solution at a 600 m 3 /h flow-rate. If conversion of HNO 3 into NO, NO 2 , at the enriching end of the columns, would be done with gaseous SO 2 , for an industrial plant of 100 t/y 15 N a consumption of 4 million t SO 2 /y and a production of 70 % H 2 SO 4 waste solution of 4.5 million m 3 /y are estimated. The reconversion of H 2 SO 4 into SO 2 in order to recycle SO 2 is a problem to be solved to compensate the cost of sulfur dioxide and to diminish the amount of sulfuric acid waste solution. It should be taken into consideration an important price reduction of 15 N in order to make possible its utilization for industrial production of nitride type nuclear fuel. (authors)

  2. Development of an advanced 16x165 Westinghouse type PWR fuel assembly for Slovenia

    International Nuclear Information System (INIS)

    Boone, M. L.; King, S. J.; Pulver, E. F.; Jeon, K.-L.; Esteves, R.; Kurincic, B.

    2004-01-01

    Industrias Nucleares do Brasil (INB), KEPCO Nuclear Fuel Company, Ltd. (KNFC), and Westinghouse Electric Company (Westinghouse) have jointly designed an advanced 16x16 Westinghouse type PWR fuel assembly. This advanced 16x16 Westinghouse type PWR fuel assembly, which will be implemented in both Kori Unit 2 (in Korea) and Angra Unit 1 (in Brazil) in January and March 2005, respectively, is an integral part of the utilities fuel management strategy. This same fuel design has also been developed for future use in Krsko Unit 1 (in Slovenia). In this paper we will describe the front-end nuclear fuel management activities utilized by the joint development team and describe how these activities played an integral part in defining the direction of the advanced 16x16 Westinghouse type PWR fuel assembly design. Additionally, this paper will describe how this design demonstrates improved margins under high duty plant operating conditions. The major reason for initiating this joint development program was to update the current 16x16 fuel assembly, which is also called 16STD. The current 16STD fuel assembly contains a non-optimized fuel rod diameter for the fuel rod pitch (i.e. 9.5 mm OD fuel rods at a 0.485 inch pitch), non-neutronic efficient components (i.e. Inconel Mid grids), no Intermediate Flow Mixer (IFM) grids, and other mechanical features. The advanced 16x16 fuel assembly is being designed for peak rod average burnups of up to 75 MWd/kgU and will use an optimized fuel rod diameter (i.e. 9.14 mm OD ZIRLO TM fuel rods), neutronic efficient components (i.e. ZIRLO TM Mid grids), ZIRLO TM Intermediate Flow Mixer (IFM) grids to improve Departure from Nucleate Boiling (DNB) margin, and many other mechanical features that improve design margins. Nuclear design activities in the areas of fuel cycle cost and fuel management were performed in parallel to the fuel assembly design efforts. As the change in reactivity due to the change in the fuel rod diameter influences directly

  3. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    Kato, Shigeru.

    1993-01-01

    In the fuel assembly of the present invention, a means for mounting and securing short fuel rods is improved. Not only long fuel rods but also short fuel rods are disposed in channel of the fuel assembly to improve reactor safety. The short fuel rods are supported by a screw means only at the lower end plug. The present invention prevents the support for the short fuel rod from being unreliable due to the slack of the screw by the pressure of inflowing coolants. That is, coolant abutting portions such as protrusions or concave grooves are disposed at a portion in the channel box where coolants flowing from the lower tie plate, as an uprising stream, cause collision. With such a constitution, a component caused by the pressure of the flowing coolants is formed. The component acts as a rotational moment in the direction of screwing the male threads of the short fuel rod into the end plug screw hole. Accordingly, the screw is not slackened, and the short fuel rods are mounted and secured certainly. (I.S.)

  4. Fuel and nuclear fuel cycle

    International Nuclear Information System (INIS)

    Prunier, C.

    1998-01-01

    The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

  5. Fuels planning: science synthesis and integration; social issues fact sheet 02: Developing personal responsibility for fuels reduction: Types of information to encourage proactive behavior

    Science.gov (United States)

    Rocky Mountain Research Station USDA Forest Service

    2004-01-01

    Fuels management responsibilities may include providing local property owners with the information for taking responsibility for reducing fuels on their land. This fact sheet discusses three different types of information that may be useful in programs to engage property owners in fuel reduction activities.

  6. Development of a 200kW multi-fuel type PAFC power plant

    Energy Technology Data Exchange (ETDEWEB)

    Take, Tetsuo; Kuwata, Yutaka; Adachi, Masahito; Ogata, Tsutomu [NTT Integrated Information & Energy System Labs., Tokyo (Japan)

    1996-12-31

    Nippon Telegraph and Telephone Corporation (NFT) has been developing a 200 kW multi-fuel type PAFC power plant which can generate AC 200 kW of constant power by switching fuel from pipeline town gas to liquefied propane gas (LPG) and vice versa. This paper describes the outline of the demonstration test plant and test results of its fundamental characteristics.

  7. Investigation and proposal of the system to affect nuclear fuel type authorization and analysis code certification

    International Nuclear Information System (INIS)

    2006-03-01

    In order to develop the system to affect more advanced and rational regulations of nuclear fuels and earlier introduction of new technologies in nuclear power plants, domestic and overseas safety regulation systems and state of their implementation for water cooled reactor fuel and safety analysis code had been investigated and new regulation system to affect nuclear fuel type authorization and analysis code certification was proposed. Topical report system for common parts related with nuclear fuel type authorization and analysis code certification was firstly proposed for knowledge base. Maintaining consistent safety examination supported by experts, introduction of domestic efficient system for lead irradiation test fuel, and analysis code certification and quality assurance were also proposed. (T. Tanaka)

  8. Studying some regimes of the WWER-440 type reactor failed fuel element operation

    International Nuclear Information System (INIS)

    Aksenov, N.A.; Samsonov, B.V.; Sulaberidze, V.Sh.; Frej, A.K.

    1981-01-01

    The results of investigating the serviceability of experimental fuel elements close by type to that of the WWER-440 type reactor in the cans of which untightness in the form of small opening are made. The tests are carried out in the SM-2 reactor high temperature water loop at the temperature of 473 K, pressure of (1-2)x10 4 kPa, coolant flow rate of 3.7-5.5 m 3 /h. The analysis of the obtained results shows that the character of changes in the fission product (FP) activity in the circuit in a considerable extent is determined bt the thermal-optical conditions of the fuel element operation. If water in the gap between fuel and can does not boil, activity changes smoothly and bursts caused by increased FP release are observed only under transient conditions of reactor operation. In the presence of water boiling in the gap the FP release has of impulse character with the frequency determined besides the untightness dimension by free volume inside the fuel element can (with its increase the pulsation frequency increases). FP release from fuel is connected with their direct escape from an open surface. When water in the gap the FP release from the fuel element occurs practically immediately. Without boiling the FP delay in the gap is determined by their diffusion in a layer of water. The conclusion is drawn that the FP release from failed fuel elements may be reduced by eliminating the water boiling in the gap between the fuel and the can by means of the fuel element power or coolant temperature decrease

  9. Calculation analysis of TRIGA MARK II reactor core composed of two types of fuel elements

    International Nuclear Information System (INIS)

    Ravnik, M.

    1988-11-01

    The most important properties of mixed cores are treated for TRIGA MARK II reactor, composed of standard (20% enriched, 8.5w% U content) and FLIP (70% enriched, 8.5w% U content) fuel elements. Large difference in enrichment and presence of burnable poison in FLIP fuel have strong influence on the main core characteristics, such as: fuel temperature coefficient, power defect, Xe and Sm worth, power and flux distributions, etc. They are significantly different for both types of fuel. Optimal loading of mixed cores therefore strongly depends on the loading pattern of both types of fuel elements. Results of systematic calculational analysis of mixed cores are presented. Calculations on the level of fuel element are performed with WIMSD-4 computer code with extended cross-section library. Core calculations are performed with TRIGAP two-group 1-D diffusion code. Results are compared to measurements and physical explanation is provided. Special concern is devoted to realistic mixed cores, for which optimal in-core fuel management is derived. Refs, figs and tabs

  10. Rearrangement of fuel assemblies in the RBMK type reactors to flatten power distribution and improve the fuel cycle

    International Nuclear Information System (INIS)

    Mityaev, Yu.I.; Vikulov, V.K.

    1982-01-01

    A possibility of increasing the burnup of uranium fuel unloaded from the RBMK type reactors is investigated. Three variants of a two-zone reactor-refueling are considered: 1. the simplest variant of continuous refueling used at present, when the central and peripherical reactor zones are additionally fueled independently by similar fuel assemblies (FA); 2. the variant under which new FA are loaded to the peripherical zone and are used there up to the same burnup as in the first case, then all the peripherical FA (PFA) are rearranged to the centre and they are used there up to maximum burnup; 3. the same as in the second variant, but not all the PFA are rearranged to the centre but only FA with small fuel burnup. It is shown by calculation that average fuel burnup for the third refueling variant is several per cent higher at the optimal burnup of rearranged FA. Besides, flattening of fuel channel power is improved in this case, that permits to increase uranium enrichment and burnup at the same maximum power. It essentially improves economic parameters of the reactor. It is concluded that realization of the considered variant of fuel refueling will produce the most essential effect for reactors refueled without shutdown

  11. Radiological impact of plutonium recycle in the fuel cycle of LWR type reactors: professional exposure during mormal operation

    International Nuclear Information System (INIS)

    White, I.F.; Kelly, G.N.

    1983-01-01

    The radiological impact of the fuel cycle of light water type reactors using enriched uranium may be changed by plutonium recycle. The impact on human population and on the persons professionally exposed may be different according to the different steps of the fuel cycle. This report analyses the differential radiological impact on the different types of personnel involed in the fuel cycle. Each step of the fuel cycle is separately studied (fuel fabrication, reactor operation, fuel reprocessing), as also the transport of the radioactive materials between the different steps. For the whole fuel cycle, one estimates that, with regard to the fuel cycle using enriched uranium, the plutonium recycle involves a small increase of the professional exposure

  12. Irradiation tests on PHWR type fuel elements in TRIGA research reactor of INR Pitesti

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore [Institute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.; Sorescu, Ion [Institute for Nuclear Research, Pitesti (Romania). TRIGA Reactor Loop Facility; Parvan, Marcel [Institute for Nuclear Research, Pitesti (Romania). Hot Cells Lab.

    2012-12-15

    Nine PHWR type fuel elements with reduced length were irradiated in loop A of the TRIGA Research Reactor of INR Pitesti. The primary objective of the test was to determine the performance of nuclear fuel fabricated at INR Pitesti at high linear powers in pressurized water conditions. Six fuel elements were irradiated with a ramp power history, achieving a maximum power of 45 kW/m during pre-ramp and of 64 kW/m in the ramp. The maximum discharge burnup was of 216 MWh/kgU. Another three fuel elements with reduced length were irradiated with declining power history. At the beginning of irradiation the fuel elements achieved a maximum linear power of 66 kW/m. The maximum fuel power was about 1.3 times the maximum expected in PHWR. The maximum discharge burnup was 205 MWh/kgU. The elements were destructively examined in the hot cells of INR Pitesti. Temperature-sensitive parameters such as UO{sub 2} grain growth, fission-gas release and sheath deformations were examined. The tests proved the feasibility of irradiating PHWR type fuel elements at linear powers up to 66 kW/m under pressurized water conditions and demonstrated the possibility of more flexible operation of this fuel in power reactors. This paper presents the results of the investigation. (orig.)

  13. CANDU type fuel activities in Argentina

    International Nuclear Information System (INIS)

    Lavarez, L.; Casario, J.A.; Moreno, C.

    2003-01-01

    Domestic fuel performance in Embalse NPP during the last two years has been excellent without a significant occurrence of fuel failures. The defect rate level was reasonably low with a lowest value of 0.02 % in 2002. The implementation of fuel design optimizations to increase uranium content was fully completed by the end of year 2000. The in-reactor performance was not affected and shows the high degree of maturity reached for both the design and the manufacturing procedures and capabilities. A feasibility study for the utilization of SEU in Embalse NPP mainly conducted by NA-SA and AECL is almost completed. Some fuel related activities are still in progress. As part of them fuel behavior simulations using simplified power histories were performed to assess the influence of SEU fuel burnup extension. (author)

  14. Contact-type displacement measuring mechanism for fuel assembly in reactor

    International Nuclear Information System (INIS)

    Yokota, Yoshio; Ko, Kuniaki.

    1995-01-01

    The measuring mechanism of the present invention, which is used in a lmfbr type reactor, is suspended by a gripper of a fuel handing machine, and it comprises a combination of a displacement amount measuring jig allowed to be inserted into a handling head of a fuel assembly and a displacement amount measuring ring disposed at the lower portion in the handling head. The displacement amount measuring jig has a structure comprising a releasable handle and a columnar or cylindrical measuring portion allowable to be inserted into the handling head formed at the lower portion of the handle, which are connected with each other. When an interference (contact) occurred between the displacement amount measuring jig and the stepwise displacement amount measuring ring during the measurement, change of load and a phenomenon that the fuel handing machine can not be lowered are recognized, so that core displacement amount can be recognized based on the stroke of the gripper portion. Then, remote measurement is possible for displacement and deformation of the fuel assembly in the reactor container, and the measurement can be conducted by the same procedures and in the same period of time as in a case of ordinary fuel exchange operation. A flow channel for coolants passing through the fuel assembly can be ensured, thereby enabling to measure the amount of core displacement which is closer to an actual value in the reactor. (N.H.)

  15. A review of microstructural analysis on U3Si2-Al plate-type fuel

    International Nuclear Information System (INIS)

    Ti Zhongxin; Guo Yibai

    1995-12-01

    The microstructure of U 3 Si 2 -Al plate-type fuel, that is the microstructure of fuel particles, compatibility of the fuel particles and Al matrix, fuel particles distribution, dogbone area morphology, clad and meat thickness, bone quality of clad/frame and clad/fuel core, and the effect of these factors on products quality were comprehensively investigated and analyzed by means of optical microscopy (OM), scanning electron microscopy (SEM), X-ray diffraction (XRD), energy dispersive X-ray spectrometry (EDX), image processing technique, etc.. The main results are as following: U-7.7%Si alloy contains two phases: primary U 3 Si 2 and small amount of USi (about 12%), free-uranium was not detected in fuel particles; the dogbone area is the key factor affecting fuel plate quality (1 ref., 16 figs., 4 tabs.)

  16. Prediction for the flow distribution and the pressure drop of a plate type fuel assembly

    International Nuclear Information System (INIS)

    Park, Jong Hark; Jo, Dea Sung; Chae, Hee Taek; Lee, Byung Chul

    2011-01-01

    A plate type fuel assembly widely used in many research reactors does not allow the coolant to mix with neighboring fuel channels due to the completely separated flow channels. If there is a serious inequality of coolant distribution among channels, it can reduce thermal-hydraulic safety margin, as well as it can cause a deformation of fuel plates by the pressure difference between neighboring channels, thus the flow uniformity in the fuel assembly should be confirmed. When designing a primary cooling system (PCS), the pressure drop through a reactor core is a dominant value to determine the PCS pump size. The major portion of reactor core pressure drop is caused by the fuel assemblies. However it is not easy to get a reasonable estimation of pressure drop due to the geometric complexity of the fuel assembly and the thin gaps between fuel assemblies. The flow rate through the gap is important part to determine the total flow rate of PCS, so it should be estimated as reasonable as possible. It requires complex and difficult jobs to get useful data. In this study CFD analysis to predict the flow distribution and the pressure drop were conducted on the plate type fuel assembly, which results would be used to be preliminary data to determine the PCS flow rate and to improve the design of a fuel assembly

  17. Design and analytic evaluation of a rim effect reduction type LWR fuel for extending burnup

    International Nuclear Information System (INIS)

    Matsumura, Tetsuo; Kameyama, Takanori; Kinoshita, Motoyasu

    1991-01-01

    We have designed a new concept fuel design 'Rim effect reduction type fuel' which has thin natural UO 2 layer on surface of a UO2 pellet. Our neutronic analyses with ANRB code show this fuel design can reduce rim effect (burnup at plelet rim) by about 30 GWd/t comparing a normal fuel. It is known that a high burnup fuel has different microstructure from as-fabricated one at fuel rim (which is called as rim region) due to rim effect. Therefore this fuel design can expect smaller rim region than a normal fuel. Our fuel performance analyses with EIMUS code show this fuel design can reduce fuel center temperature at high burnup if thermal conductivity of fuel pellet decreases with burnup in inverse proportion. However, this fuel design increases fuel center temperature at low and middle burnup than a normal fuel due to increase of thermal power density at pellet center. Additionally Irradiation experiment of this fuel design can be considered to offer important data which make clear the relation between rim effect and fuel performance. (author)

  18. Analysis on small long life reactor using thorium fuel for water cooled and metal cooled reactor types

    International Nuclear Information System (INIS)

    Permana, Sidik

    2009-01-01

    Long-life reactor operation can be adopted for some special purposes which have been proposed by IAEA as the small and medium reactor (SMR) program. Thermal reactor and fast reactor types can be used for SMR and in addition to that program the utilization of thorium fuel as one of the candidate as a 'partner' fuel with uranium fuel which can be considered for optimizing the nuclear fuel utilization as well as recycling spent fuel. Fissile U-233 as the main fissile material for thorium fuel shows higher eta-value for wider energy range compared with other fissile materials of U-235 and Pu-239. However, it less than Pu-239 for fast energy region, but it still shows high eta-value. This eta-value gives the reactor has higher capability for obtaining breeding condition or high conversion capability. In the present study, the comparative analysis on small long life reactor fueled by thorium for different reactor types (water cooled and metal cooled reactor types). Light water and heavy water have been used as representative of water-cooled reactor types, and for liquid metal-cooled reactor types, sodium-cooled and lead-bismuth-cooled have been adopted. Core blanket arrangement as general design configuration, has been adopted which consist of inner blanket region fueled by thorium oxide, and two core regions (inner and out regions) fueled by fissile U-233 and thorium oxide with different percentages of fissile content. SRAC-CITATION and JENDL-33 have been used as core optimization analysis and nuclear data library for this analysis. Reactor operation time can reaches more than 10 years operation without refueling and shuffling for different reactor types and several power outputs. As can be expected, liquid metal cooled reactor types can be used more effective for obtaining long life reactor with higher burnup, higher power density, higher breeding capability and lower excess reactivity compared with water-cooled reactors. Water cooled obtains long life core operation

  19. Fuel exchange device for FBR type reactor

    International Nuclear Information System (INIS)

    Onuki, Koji.

    1993-01-01

    The device of the present invention can provide fresh fuels with a rotational angle aligned with the direction in the reactor core, so that the fresh fuels can be inserted being aligned with apertures of the reactor core even if a self orientation mechanism should fail to operate. That is, a rotational angle detection means (1) detects the rotational angle of fresh fuels before insertion to the reactor core. A fuel rotational angle control means (2) controls the rotational angle of the fresh fuels by comparing the detection result of the means (1) and the data for the insertion position of the reactor core. A fuel rotation means (3) compensates the rotational angel of the fresh fuels based on the control signal from the means (2). In this way, when the fresh fuels are inserted to the reactor core, the fresh fuels set at the same angle as that for the aperture of the reactor core. Accordingly, even if the self orientation mechanism should not operate, the fresh fuels can be inserted smoothly. As a result, it is possible to save loss time upon fuel exchange and mitigate operator's burden during operation. (I.S.)

  20. Fuel cladding tube and fuel rod for BWR type reactor

    International Nuclear Information System (INIS)

    Urata, Megumu; Mitani, Shinji.

    1995-01-01

    A fuel cladding tube has grooves fabricated, on the surface thereof, with a predetermined difference between crest and bottom (depth of the groove) in the circumferential direction. The cross sectional shape thereof is sinusoidal. The distribution of the grain size of iron crud particles in coolants is within a range about from 2μm to 12μm. If the surface roughness of the fuel cladding tube (depth of the groove) is determined greater than 1.6μm and less than 12.5, iron cruds in coolants can be positively deposited on the surface of the fuel cladding tube. In addition, once deposited iron cruds can be prevented from peeling from the surface of the fuel cladding tube. With such procedures, iron cruds deposited and radioactivated on the fuel cladding tube can be prevented from peeling, to prevent and reduce the increase of radiation dose on the surface of the pipelines without providing any additional device. (I.N.)

  1. Loss-of-flow test L5 on FFTF-type irradiated fuel

    International Nuclear Information System (INIS)

    Simms, R.; Gehl, S.M.; Lo, R.K.; Rothman, A.B.

    1978-03-01

    Test L5 simulated a hypothetical loss-of-flow accident in an LMFBR using three (Pu, U)O 2 fuel elements of the FTR type. The test elements were irradiated before TREAT Test L5 in the General Electric Test Reactor to 8 at. % burnup at about 40 kW/m. The preirradiation in GETR caused a fuel-restructuring range characteristic of moderate-power structure relative to the FTR. The test transient was devised so that a power burst would be initiated at incipient cladding melting after the loss of flow. The test simulation corresponds to a scenario for FTR in which fuel in high-power-structure subassemblies slump, resulting in a power excursion. The remaining subassemblies are subjected to this power burst. Test L5 addressed the fuel-motion behavior of the subassemblies in this latter category. Data from test-vehicle sensors, hodoscope, and post-mortem examinations were used to construct the sequence of events within the test zone. From these observations, the fuel underwent a predominantly dispersive event just after reaching a peak power six times nominal at, or after, scram. The fuel motion was apparently driven by the release of entrained fission-product gases, since fuel vapor pressure was deliberately kept below significant levels for the transient. The test remains show a wide range of microstructural evolution, depending on the extent of heat deposition along the active fuel column. Extensive fuel swelling was also observed as a result of the lack of the cladding restraint. The results of the thermal-hydraulic calculations with the SAS3A code agreed qualitatively with the postmortem results with respect to the extent of the melting and the dispersal of cladding and fuel. However, the calculated times of certain events did not agree with the observed times

  2. Porosity in MX-type fuels and its stability

    International Nuclear Information System (INIS)

    Sari, C.

    1978-01-01

    Radial and axial temperature gradients were generated in MX-type fuels (U,Pu)C, (U,Pu)CN and (U,Pu)N in regions of temperature between 1000 and 2000 0 C. Typical temperature gradients were between 150 and 350 0 C/mm. Experiments show that under these conditions important restructuring of the fuel occurs after less than 40 hours. Densification in the thermal gradient was observed at temperature as low as 1100 0 C and the densification decreases with the increase of the nitrogen content. The grain growth rates decrease with the increase of the nitrogen content, thus paralleling the results of densification. Evidence of pore migration was found in the region with T approximately equal to 1500 0 C. Data of pore migration in MC and in carbon rich MCN plotted in an Arrhenius diagram gives a ΔH approximately equal to 95kcal/mole in approximate agreement with the values of evaporation enthalpy

  3. Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.; Billone, M.C.; Holland, J.W.; Kramer, J.M.

    1992-11-01

    This report discusses three furnace heating tests which were conducted with irradiated, HT9-clad and U-19wt.%Pu-l0wt.%Zr-alloy fuel, Mk-V-type fuel elements in the Alpha-Gamma Hot Cell Facility at Argonne National Laboratory, Illinois. In general, very significant safety margins for fuel-element cladding breaching have been demonstrated in these tests, under conditions that would envelop a bounding unlikely loss-of-flow event in EBR-II. Highlights of the test results will be given, as well as discussions of the cladding breaching mechanisms, axial fuel motion, and fuel surface liquefaction found in high-temperature testing of irradiated metallic fuel elements

  4. Improvement of the vibration of the test fuel(Type-B) with a guide tube under operational condition

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong Seung; Yim, Jeong Sik; Lim, I. C. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-04-01

    The Type-B test fuel for the Hanaro has a flexible guide tube on top of the fuel to lead and guide the instrumentation wires. Depending on the flow condition in the reactor, the fuel is susceptible to vibration. During the test operation of the fuel, a fairly large amplitude vibration was observed and the possibility of flow tube contact with adjacent flow tubes, due to the excessive vibration of the fuel, and consequent wear or defect of the flow tubes were raised. Thus, to know the vibration characteristics as well as whether the flow tube contact each other, analyses of the Type-B fuel the dummy fuel were performed by BEVIRA and ANSYS. Besides the analyses, vibration tests using the dummy fuel in air and with Type-B fuel in the core at zero power under operational flow condition were executed. The results from the analyses were compared with those from tests to validate the analyses. From the deflection test of the dummy fuel in air to get the maximum displacement of the flow tube at the top, the flow tube were found to contact each other. For the prevention of the contact of the flow tubes caused by the excessive vibration of the guide tube, an additional support to the guide tube was proposed. With the additional support, analysis and in core vibration test under operational flow condition were conducted and there found to be no excessive vibration any more. 6 refs., 16 figs., 6 tabs. (Author)

  5. Progress of the Russian RERTR program: Development of new-type fuel elements for Russian-built research reactors

    International Nuclear Information System (INIS)

    Vatulin, A. V.; Stetskiy, Y.A.; Mishunin, V.A.; Suprun, V.B.; Dobrikova, I.V.

    2002-01-01

    The new design of pin-type fuel elements and fuel assembly on their basis for Russian research reactors has been developed. The number of following activities has been performed: computational and experimental substantiation of fuel element design; development of fabrication process of fuel elements; manufacturing of experimental assembly for lifetime in-pile tests. The relevant fuel assemblies are considered to be perspective for usage as low-enriched fuel for Russian research reactors. (author)

  6. Guide for the development of power generation projects using renewable energy in and for municipalities; Guia para el desarrollo de proyectos de generacion de electricidad con energia renovable en y para los municipios

    Energy Technology Data Exchange (ETDEWEB)

    De Buen Rodriguez, Odon [Energia, Tecnologia y Educacion, S.C., ENTE, S.C. (Mexico)

    2010-05-15

    The Mexico Competitiveness Project has come about as a result of a good understanding and broad cooperation between the United States and Mexican governments. Its primary objective is to strengthen governance so as to improve the economic competitiveness of Mexico as well as to improve environmental management for the preservation of biodiversity in the country. This program, financed by the United States Agency for International Development (USAID), will be carried out for approximately 4 years, from October 2008 to November 2012. Its cooperative efforts are primarily focused on assisting local efforts to foster sustainable development. In this context, and as part of the specific activities on which the program is focused during the current year, the Guide for the Development of Power Generation Projects in Municipalities was designed, whose purpose is to integrate information to enable municipal officials to understand these types of projects and take actions to foster their development. The present guide consists of a set of modules that detail and explain the most important aspects involved in the development of these projects, ranging from legal considerations to costs associated with the projects themselves. This guide has two sections: a general portion with 8 chapters about topics directly related to the projects and a set of appendices that describe aspects relevant to moving forward with projects for power generation using renewable energy. [Spanish] El Programa de Competitividad Mexico se deriva del buen entendimiento y amplia cooperacion entre los gobiernos de Estados Unidos y Mexico, y tiene como principal objetivo fortalecer la gobernabilidad para mejorar la competitividad economica de Mexico, asi como robustecer la gestion ambiental para preservar la biodiversidad del pais. Este programa, que es financiado por la Agencia de los Estados Unidos para el Desarrollo Internacional (USAID), tiene una duracion aproximada de cuatro anos a partir de octubre

  7. Análise do uso de aplicativos para o planejamento de viagens práticas no curso Técnico em Guia de Turismo no IFRS Restinga = Analysis of the use applications for practical travel planning on Technical Tour Guide Course at IFRS Restinga

    Directory of Open Access Journals (Sweden)

    Abreu, Carina Vasconcellos

    2015-01-01

    Full Text Available O presente artigo objetiva identificar e analisar o uso de recursos tecnológicos para o planejamento de viagens práticas do curso técnico em guia de turismo do IFRS Restinga. Foram selecionados três recursos considerados mais utilizados e acessíveis aos alunos: Google Maps, GPS e Waze. O curso prevê planejamento de trajetos, condições das vias e tempo de viagem, para posterior execução de atividades práticas que simulam o trabalho do guia de turismo. Cada recurso foi descrito em suas funcionalidades, limitações e possibilidades de uso no planejamento das viagens práticas, considerando possibilidade de acesso e atualização de informações. Considera-se que todos têm potencialidade para serem utilizados, inclusive de forma combinada, objetivando a inclusão digital e a instrumentalização dos alunos para o uso dos recursos

  8. Exame físico na criança: um guia para o enfermeiro

    Directory of Open Access Journals (Sweden)

    Josilene de Melo Buriti Vasconcelos

    1999-12-01

    Full Text Available O exame físico constitui-se numa importante etapa dentro do processo de avaliação à criança, através do qual o enfermeiro enriquece as informações obtidas durante a entrevista e fundamenta a assistência de enfermagem. Na atualidade, mediante o crescente interesse dos enfermeiros em aplicar o processo de enfermagem em todas as suas etapas, o exame físico tem ocupado lugar de destaque, por permitir que os enfermeiros conheçam as necessidades de seus clientes, no que diz respeito aos seus aspectos físicos e fisiológicos, permitindo a identificação dos diagnósticos de enfermagem, alem de servir como recurso para avaliação efetiva das intervenções de enfermagem. Este estudo, de caráter bibliográfico, objetivou elaborar um guia para realização do exame físico na criança, a ser utilizado por enfermeiros, durante o processo de avaliação à criança. Nele apresentamos o exame físico de forma sistemática, utilizando a seqüência céfalo-caudal, descrevendo os possíveis achados normais e anormais ao examinarmos a criança.

  9. GUIA METODOLOGICA PARA UN MANEJO INTEGRAL COSTERO APLICADO A PEHUEN CO (ARGENTINA

    Directory of Open Access Journals (Sweden)

    María Luján Bustos

    2017-03-01

    Os planos de gestão costeira integrada (GCI na Argentina são quase inexistentes. No Sul da província de Buenos Aires (Argentina, o planejamento por causa da intensidade da erosão costeira, que afeta especialmente as áreas urbanizadas, é necessário. É fundamental ainda traçar estratégias de GCI que possam ser aplicadas em toda a área de forma sustentável. Neste processo, a participação e o compromisso dos atores sociais são fundamentais. Por esta razão, um guia metodológico foi construído para uma GCI com participação social e aplica-se à localidade de Pehuen Co. Obteve-se como resultado que a erosão costeira e a sazonalidade do turismo são os pilares tanto para a visão científica quanto para a social. Sobre isso, foram delineadas ações e estratégias a curto, médio e longo prazos e indicadores de acompanhamento. Concluiu-se que é necessário o compromisso das autoridades locais, trabalhando em conjunto com as partes interessadas, para o bom funcionamento do GCI. Palavras-chave: Gestão Integrada Costeira; Indicadores Ambientais; Erosão; Sazonalidade do Turismo; Pehuen Co (Argentina.

  10. Commuters' exposure to particulate matter air pollution is affected by mode of transport, fuel type, and route.

    Science.gov (United States)

    Zuurbier, Moniek; Hoek, Gerard; Oldenwening, Marieke; Lenters, Virissa; Meliefste, Kees; van den Hazel, Peter; Brunekreef, Bert

    2010-06-01

    Commuters are exposed to high concentrations of air pollutants, but little quantitative information is currently available on differences in exposure between different modes of transport, routes, and fuel types. The aim of our study was to assess differences in commuters' exposure to traffic-related air pollution related to transport mode, route, and fuel type. We measured particle number counts (PNCs) and concentrations of PM2.5 (particulate matter bus passengers, we calculated that the inhaled air pollution doses were highest for cyclists. With the exception of PM10, we found that inhaled air pollution doses were lowest for electric bus passengers. Commuters' rush hour exposures were significantly influenced by mode of transport, route, and fuel type.

  11. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    Ueda, Makoto

    1990-01-01

    Various considerations are applied to fuel rods for improving the fuel burnup degree. If a gap between the fuel rods is changed, this varies the easiness for the flow of coolants depending on places, to reduce the thermal margin. Then, it is noted for the distribution of stresses generated due to the difference of water pressure caused by the difference of water streams between the inside and the outside of a channel box, and composite value, of stresses upon occurrence of earthquakes, neutron irradiation and a channel creep phenomenon caused by the stresses of due to the water pressure difference described above, the thickness of the channel box is increased in the upstream and decreased toward the downstream. Further, fuel spacers at the position where the thickness of the channel box is changed are spaced apart from the channel box so as not to brought into contact with the channel box. This can contribute to the reduction of coolants pressure loss, improvement of critical power and improvement of reactivity, as well as remarkably moderate local stresses applied from the fuel spacers to the channel box due to horizontal vibrations upon occurrence of earthquakes to improve the integrity of fuel assembly. (N.H.)

  12. Major results on the development of high density U-Mo fuel and pin-type fuel elements executed under the Russian RERTR program and in cooperation with ANL (USA)

    International Nuclear Information System (INIS)

    Vatulin, A.; Morozov, A.; Stetsky, Y.; Suprun, V.; Dobrikova, I.; Trifonov, Y.; Mishunin, V.; Sorokin, V.

    2003-01-01

    VNIINM is active participant of 'Russian program on Reduced Enrichment for Research and Test Reactors'. Institute Works in two main directions: 1) development of new high-density fuels (HDF) and 2) development of new design of fuel elements with LEU. The development of the new type fuel element is carried out both for existing reactors, and for developing new advanced reactors. The 'TVEL' concern is coordinator of works of this program. The majority enterprises of branch (NIIAR, PIYaF, RRC KI, NZChK) take part in this work. Since 2000 these works are being conducted in cooperation with Argonne National Laboratory (USA) within the RERTR program under VNIINM with ANL contract. At the present, a large set of pre-pile investigations has been completed. All necessary fabrication procedures have been developed for utilization of U-Mo dispersion fuel in Russian-designed research reactors. For irradiation tests the pin-type mini-fuel elements with HDF dispersion fuel with LEU and the uranium density equaled to 4,0 and 6,0 g/cm 3 (up to 40 vol.%) have been manufactured. Their irradiation began in August 2003 in the MIR reactor (NIIAR, Dimitrovgrad). A large set of works for preparation of lifetime tests (WWR-M reactor in Gatchina) of two full-scale fuel assemblies with new pin-type fuel elements on basis LEU UO 2 -Al and UMo-Al fuels has been completed. The in-pile tests of fuel assemblies began in September 2003. The summary of important results of performed works and their near-term future are presented in paper. (author)

  13. Thermally induced dispersion mechanisms for aluminum-based plate-type fuels under rapid transient energy deposition

    International Nuclear Information System (INIS)

    Georgevich, V.; Taleyarkham, R.P.; Navarro-Valenti, S.; Kim, S.H.

    1995-01-01

    A thermally induced dispersion model was developed to analyze for dispersive potential and determine onset of fuel plate dispersion for Al-based research and test reactor fuels. Effect of rapid energy deposition in a fuel plate was simulated. Several data types for Al-based fuels tested in the Nuclear Safety Research Reactor in Japan and in the Transient Reactor Test in Idaho were reviewed. Analyses of experiments show that onset of fuel dispersion is linked to a sharp rise in predicted strain rate, which futher coincides with onset of Al vaporization. Analysis also shows that Al oxidation and exothermal chemical reaction between the fuel and Al can significantly affect the energy deposition characteristics, and therefore dispersion onset connected with Al vaporization, and affect onset of vaporization

  14. Utilization of radiographic and ultrasonic testing for an evaluation of plate type fuel elements during manufacturing stages

    International Nuclear Information System (INIS)

    Brito, Mucio Jose Drummond de; Silva Junior, Silverio Ferreira da; Messias, Jose Marcos; Braga, Daniel Martins; Paula, Joao Bosco de

    2005-01-01

    Structural discontinuities can be introduced in the plate type fuel elements during the manufacturing stages due to mechanical processing conditions. The use of nondestructive testing methods to monitoring the fuel elements during the manufacturing stages presents a significant importance, contributing for manufacturing process improvement and cost reducing. This paper describes a procedure to be used detection and evaluation of structural discontinuities in plate type fuel elements during the manufacturing stages using the ultrasonic testing method and the radiographic testing method. The main results obtained are presented and discussed. (author)

  15. On the possibility of reprocessing of fuel elements of dispersion type with copper matrix by pyrochemical methods

    International Nuclear Information System (INIS)

    Vasin, B.D.; Ivanov, V.A.; Shchetinskij, A.V.; Vavilov, S.K.; Savochkin, Yu.P.; Bychkov, A.V.; Kormilitsyn, M.V.

    2005-01-01

    A consideration is given to pyrochemical processes suitable for separation of uranium dioxide from structural materials when reprocessing cermet type fuel elements. The estimation of the possibility to apply liquid antimony and bismuth, potassium and copper chlorides melts is made. The specimens compacted of copper and uranium dioxide powders in a stainless steel can are used as simulators of fuel element sections. It is concluded that the dissolution of structural materials in molten salts at the stage of uranium dioxide concentration is the process of choice for reprocessing of dispersion type fuel elements [ru

  16. Conceptual design of control rod regulating system for plate type fuels of Triga-2000 reactor

    International Nuclear Information System (INIS)

    Eko Priyono; Saminto

    2016-01-01

    Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor has been made. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor was made with refer to study result of instrument and control system which is used in BATAN'S reactor. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor consist of 4 segments that is control panel, translator, driver and display. Control panel is used for regulating, safety and display control rod, translator is used for signal processing from control panel, driver is used for driving control rod and display is used for display control rod level position. The translator was designed in 2 modes operation i.e operation by using PLC modules and IC TTL modules. These conceptual design can be used as one of reference of control rod regulating system detail design. (author)

  17. The influence of the types of marine fuel over the Energy Efficiency Operational Index

    Science.gov (United States)

    Acomi, Nicoleta; Acomi, Ovidiu

    2014-05-01

    One of the main concerns of our society is certainly the environment protection. The international efforts for maintaining the environment clean are various and this paper refers to the efforts in the maritime transport field. Marine pollution consists of the water pollution and also the air pollution. Regardless of the delay in recognizing the later type of pollution, it rapidly gains many organizations to argue on it. The first step was including a dedicated annex (Annex VI) in the International Convention for the Prevention of Pollution from Ships, in 1997, which seeks to minimize the airborne emissions from ships. In order to control and minimize the air pollution, the International Maritime Organization has also developed a series of measures for monitoring the emissions. These measures are grouped in three main directions: technical, operational and management related. The subject of our study is the concept of Energy Efficiency Operational Index (EEOI), developed to provide ship-owners with assistance in the process of establishing the emissions from ships in operation, and to suggest the methods for achieving their reduction. As a monitoring tool, EEOI represents the mass of CO2 emitted per unit of transport work. The actual CO2 emission from combustion of fuel on board a ship during each voyage is calculated by multiplying total fuel consumption for each type of fuel (e.g. diesel oil, gas oil, light fuel oil, heavy fuel oil, liquefied petroleum gas, liquefied natural gas) with the carbon to CO2 conversion factor for the fuel in question. The performed transport work is calculated by multiplying mass of cargo (tonnes, number of TEU/cars, or number of passengers) with the distance in nautical miles corresponding to the transport work done. Using the software developed by the author it will be emphasized the variation of the EEOI value for one vessel using different types of fuel for the voyage's legs (distance to discharge port, distance to loading port, the

  18. Fission rate measurements in fuel plate type assembly reactor cores

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1988-01-01

    The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs

  19. Taguchi Method for Investigating the Performance Parameters and Exergy of a Diesel Engine Using Four Types of Diesel Fuels

    Directory of Open Access Journals (Sweden)

    Dara K. Khidir

    2016-05-01

    Full Text Available The effects of changes in engine operating parameters, i.e., engine speed, throttle and water temperature, for four types of diesel fuel (A, B, C and D of different specific gravities, as supplied from local market and refineries, were studied and simultaneously optimized. The experiment design was based on Taguchi’s “L' 16” orthogonal table, and the engine was put to test at different engine speeds, throttling opening percentages and water temperatures, using different fuels. The data were analyzed using S/N (signal to noise ratio for each factor. The obtained results show that the optimum operating conditions for minimum BSFC (brake specific fuel consumption are achieved when the engine speed is 2500 rpm, the throttle is placed at 75% of full throttling, the water temperature is 80 oC and the engine is using fuel type D. Also, results of S/N ratio reveal that the throttle has significant influence on brake thermal and exergic efficiencies. Water temperature is the second most effective factor and then comes the influence of engine speed. The least effective factor among the studied parameters for the types of fuel considered in this experiment is the fuel type.

  20. Burn-up calculations for a thorium HTR with one and with two types of fuel particle

    Energy Technology Data Exchange (ETDEWEB)

    Griggs, C. F.

    1975-06-15

    Cell burn-up calculations have been made on a thorium pin-cell operating with one or with two types of particle. With one particle, the input thorium and uranium are mixed prior to irradiation and all discharged uranium is recycled. With two particles, the fuel is kept in two streams and only the uranium generated from thorium is recycled. The two models are found to give similar power generations from a given initial U-235 input. The choice between the two types of particle is probably not determined by reactor physics considerations but by the value of the fuel credits and by the cost of fuel fabrication and reprocessing.

  1. Artificial intelligence applied to fuel management in BWR type reactors

    International Nuclear Information System (INIS)

    Ortiz S, J.J.

    1998-01-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  2. Technical report: technical development on the silicide plate-type fuel experiment at nuclear safety research reactor

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Soyama, Kazuhiko; Ichikawa, Hiroki

    1991-08-01

    According to a reduction of fuel enrichment from 45 w/o 235 U to 20 w/o, an aluminide plate-type fuel used currently in the domestic research and material testing reactors will be replaced by a silicide plate-type one. One of the major concern arisen from this alternation is to understand the fuel behavior under simulated reactivity initiated accident (RIA) conditions, this is strongly necessary from the safety and licensing point of view. The in-core RIA experiments are, therefore, carried out at Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Research Institute (JAERI). The silicide plate-type fuel consisted of the ternary alloy of U-Al-Si as a meat with uranium density up to 4.8 g/cm 3 having thickness by 0.51 mm and the binary alloy of Al-3%Mg as a cladding by thickness of 0.38 mm. Comparison of the physical properties of this metallic plate fuel with the UO 2 -zircaloy fuel rod used conventionally in commercial light water reactors shows that the heat conductivity of the former is of the order of about 13 times greater than the latter, however the melting temperature is only one-half (1570degC). Prior to in-core RIA experiments, there were some difficulties lay in our technical path. This report summarized the technical achievements obtained through our four years work. (J.P.N.)

  3. Effect of engine parameters and type of gaseous fuel on the performance of dual-fuel gas diesel engines. A critical review

    Energy Technology Data Exchange (ETDEWEB)

    Sahoo, B.B. [Centre for Energy, Indian Institute of Technology, Guwahati 781039 (India); Sahoo, N.; Saha, U.K. [Department of Mechanical Engineering, Indian Institute of Technology, Guwahati 781039 (India)

    2009-08-15

    Petroleum resources are finite and, therefore, search for their alternative non-petroleum fuels for internal combustion engines is continuing all over the world. Moreover gases emitted by petroleum fuel driven vehicles have an adverse effect on the environment and human health. There is universal acceptance of the need to reduce such emissions. Towards this, scientists have proposed various solutions for diesel engines, one of which is the use of gaseous fuels as a supplement for liquid diesel fuel. These engines, which use conventional diesel fuel and gaseous fuel, are referred to as 'dual-fuel engines'. Natural gas and bio-derived gas appear more attractive alternative fuels for dual-fuel engines in view of their friendly environmental nature. In the gas-fumigated dual-fuel engine, the primary fuel is mixed outside the cylinder before it is inducted into the cylinder. A pilot quantity of liquid fuel is injected towards the end of the compression stroke to initiate combustion. When considering a gaseous fuel for use in existing diesel engines, a number of issues which include, the effects of engine operating and design parameters, and type of gaseous fuel, on the performance of the dual-fuel engines, are important. This paper reviews the research on above issues carried out by various scientists in different diesel engines. This paper touches upon performance, combustion and emission characteristics of dual-fuel engines which use natural gas, biogas, producer gas, methane, liquefied petroleum gas, propane, etc. as gaseous fuel. It reveals that 'dual-fuel concept' is a promising technique for controlling both NO{sub x} and soot emissions even on existing diesel engine. But, HC, CO emissions and 'bsfc' are higher for part load gas diesel engine operations. Thermal efficiency of dual-fuel engines improve either with increased engine speed, or with advanced injection timings, or with increased amount of pilot fuel. The ignition

  4. Management of the acceptance process of RTR aluminide type spent fuel

    International Nuclear Information System (INIS)

    Auziere, P.; Thomasson, J.

    2002-01-01

    A wide range of Research Test Reactor aluminide type spent fuel is already received for treatment conditioning at the La Hague reprocessing complex. Such a diversity calls for an utmost attention to be paid to all safety-related systems and technical aspects, to all regulatory and administrative constraints. Despite of such multiple data inputs and rigid constraints, a close cooperation between the Research Reactor operator and COGEMA enables to reach adequate and cost effective solutions also relevant to spent fuel having had an uneven history. The acceptance process is primarily based on the client descriptive data and status declaration issued by the Research Reactor (RR) operator under QA. This acceptance process is a key step, to be keenly scheduled as it is directly interactive with the RR evacuation plans and the La Hague industrial plant program. It is also governed by the reviews conducted by the French Safety Authority and generally translated into operational authorisations. Concerned by maintaining high safety standards, reliable and proven operational levels of its nuclear services performed in the La Hague facilities COGEMA includes, all through this acceptance process, the operating, regulatory and administrative requirements. This paper sets forth an overview of the approach implemented in the COGEMA organisation for the management of the acceptance process of RTR aluminide type spent fuel. (author)

  5. Design of an equilibrium nucleus of a BWR type reactor based in a Thorium-Uranium fuel

    International Nuclear Information System (INIS)

    Francois, J.L.; Nunez C, A.

    2003-01-01

    In this work the design of the reactor nucleus of boiling water using fuel of thorium-uranium is presented. Starting from an integral concept based in a type cover-seed assemble is carried out the design of an equilibrium reload for the nucleus of a reactor like that of the Laguna Verde Central and its are analyzed some of the main design variables like the cycle length, the reload fraction, the burnt fuel, the vacuum distribution, the generation of lineal heat, the margin of shutdown, as well as a first estimation of the fuel cost. The results show that it is feasible to obtain an equilibrium reload, comparable to those that are carried out in the Laguna Verde reactors, with a good behavior of those analyzed variables. The cost of the equilibrium reload designed with the thorium-uranium fuel is approximately 2% high that the uranium reload producing the same energy. It is concluded that it is convenient to include burnable poisons, type gadolinium, in the fuel with the end of improving the reload design, the fuel costs and the margin of shutdown. (Author)

  6. Reactor core T-H characteristics determination in case of parallel operation of different fuel assembly types

    International Nuclear Information System (INIS)

    Hermansky, J.; Petenyi, V.; Zavodsky, M.

    2009-01-01

    The WWER-440 nuclear fuel vendor permanently improve the assortment of produced nuclear fuel assemblies for achieving better fuel cycle economy and reactor operation safety. Therefore it is necessary to have the skilled methodology and computing code for analyzing factors which affecting the accuracy of flow redistributed determination through reactor on flows through separate parts of reactor core in case of parallel operation different assembly types. Whereas the geometric parameters of new manufactured assemblies were changed recently, the calculated flows through the fuel parts of different type of assemblies are depended also on their real position in reactor core. Therefore the computing code CORFLO was developed in VUJE Trnava for carrying out stationary analyses of T-H characteristics of reactor core within 60 deg symmetry. The CORFLO code deals the area of the active core which consists of 312 fuel assemblies and 37 control assemblies. Regarding the rotational 60 deg symmetry of reactor core only 1/6 of reactor core with 59 fuel assemblies is calculated. Computing code is verified and validated at this time. Paper presents the short description of computing code CORFLO with some calculated results. (Authors)

  7. Influence of the silicon content on the core corrosion properties of dispersion type fuel plates

    International Nuclear Information System (INIS)

    Calvo, C.; Saenz de Tejada, L. M.; Diaz Diaz, J.

    1969-01-01

    A new process to produce aluminium base dispersion type fuel plates has been developed at the Spanish JEN (Junta de Energia Nuclear). The dispersed fuel material is obtained by an aluminothermic process to render a stoichiometric cermet of UAI 3 and AI 2 O 3 according to the reaction. (Author)

  8. Vehicle type choice under the influence of a tax reform and rising fuel prices

    DEFF Research Database (Denmark)

    Mabit, Stefan Lindhard

    2014-01-01

    change in new vehicle purchases toward more diesel vehicles and more fuel-efficient vehicles. The paper analyses to what extent a vehicle tax reform similar to the Danish 2007 reform may explain changes in purchasing behaviour. The paper investigates the effects of a tax reform, fuel price changes......, and technological development on vehicle type choice using a mixed logit model. The model allows a simulation of the effect of car price changes that resemble those induced by the tax reform. This effect is compared to the effects of fuel price changes and technology improvements. The simulations show...... that the effect of the tax reform on fuel efficiency is similar to the effect of rising fuel prices while the effect of technological development is much larger. The conclusion is that while the tax reform appeared in the same year as a large increase in fuel efficiency, it seems likely that it only explains...

  9. Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shelley, Afroza; Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi

    2003-10-01

    Parametric studies have been performed for a PWR-type reduced-moderation water reactor (RMWR) with the seed-blanket type fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup by using MOX fuel. From the viewpoint of reactor safety analysis, the fuel temperature coefficients were also studied. From the result of the burnup calculation, it has been seen that ratio of 40-50% of outer blanket in a seed-blanket assembly gives higher conversion ratio compared to the other combination of seed-blanket assembly. And the recommended number of (seed+blanket) layers is 20, in which the number of seed (S) layers is 15 (S15) and blanket (B) layers is 5 (B5). It was found that the conversion ratio of seed-blanket assembly decreases, when they are arranged looks like a flower shape (Hanagara). By the optimization of different parameters, S15B5 fuel assembly with the height of seed of 1000 mmx2, internal blanket of 150 mm and axial blanket of 400 mmx2 is recommended for a reactor of high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and blanket fuel rod is 0.4 mm. In S15B5 assembly, the conversion ratio is 1.0 and the burnup is 38.18 GWd/t in (seed+internal blanket+outer blanket) region. However, the burnup is 57.45 GWd/t in (seed+internal blanket) region. The cycle length of the core is 16.46 effective full power in month (EFPM) by six batches and the enrichment of fissile Pu is 14.64 wt.%. The void coefficient is +21.82 pcm/%void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation. It is also possible to use S15B5 fuel assembly as a high burnup reactor 45 GWd/t in (seed+internal blanket+outer blanket) region, however, it is necessary to decrease the height of seed to 500 mmx2 to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +20.81 pcm/%void. The fuel temperature

  10. A novel reactor type for autothermal reforming of diesel fuel and kerosene

    International Nuclear Information System (INIS)

    Pasel, Joachim; Samsun, Remzi Can; Tschauder, Andreas; Peters, Ralf; Stolten, Detlef

    2015-01-01

    Highlights: • Development and experimental evaluation of Juelich’s novel ATR reactor type. • Constructive integration of steam generation chamber and nozzle for water injection. • Internal steam generator modified to reduce pressure drop to approx. a thirtieth. • Novel concept for ATR heat management proven to be suitable for fuel cell systems. • Reaction conditions during shut-down and start-up optimized to reduce byproducts. - Abstract: This paper describes the development and experimental evaluation of Juelich’s novel reactor type ATR AH2 for autothermal reforming of diesel fuel and kerosene. ATR AH2 overcomes the disadvantages of Juelich’s former reactor generations from the perspective of the fuel cell system by constructively integrating an additional pressure swirl nozzle for the injection of cold water and a steam generation chamber. As a consequence, ATR AH2 eliminates the need for external process configurations for steam supply. Additionally, the internal steam generator has been modified by increasing its cross-sectional area and by decreasing its length. This measure reduces the pressure drop of the steam generator from approx. 500 mbar to roughly a thirtieth. The experimental evaluation of ATR AH2 at steady state revealed that the novel concept for heat management applied in ATR AH2 is suitable for fuel cell systems at any reformer load point between 20% and 120% when the mass fractions of cold water to the newly integrated nozzle are set to values between 40% and 50%. The experimental evaluation of ATR AH2 during start-up and shut-down showed that slight modifications of the reaction conditions during these transient phases greatly reduced the concentrations of ethene, ethane, propene and benzene in the reformate. From the fuel cell system perspective, these improvements provide a very beneficial contribution to longer stabilities for the catalysts and adsorption materials

  11. Finite element modelling of different CANDU fuel bundle types in various refuelling conditions

    International Nuclear Information System (INIS)

    Roman, M. R.; Ionescu, D. V.; Olteanu, G.; Florea, S.; Radut, A. C.

    2016-01-01

    The objective of this paper is to develop a finite element model for static strength analysis of the CANDU standard with 37 elements fuel bundle and the SEU43 with 43 elements fuel bundle design for various refuelling conditions. The computer code, ANSYS7.1, is used to simulate the axial compression in CANDU type fuel bundles subject to hydraulic drag loads, deflection of fuel elements, stresses and displacements in the end plates. Two possible situations for the fuelling machine side stops are considered in our analyses, as follows: the last fuel bundle is supported by the two side stops and a side stop can be blocked therefore, the last fuel bundle is supported by only one side stop. The results of the analyses performed are briefly presented and also illustrated in a graphical form. The finite element model developed in present study is verified against test results for endplate displacement and element bowing obtained from strength tests with fuel bundle string and fuelling machine side-stop simulators. Comparison of ANSYS model predictions with these experimental results led to a very good agreement. Despite the difference in hydraulic load between SEU43 and CANDU standard fuel bundles strings, the maximum stress in the SEU43 endplate is about the same with the maximum stress in the CANDU standard endplate. The comparative assessment reveals that SEU43 fuel bundle is able to withstand high flow rate without showing a significant geometric instability. (authors)

  12. Study of development of non-destructive method for determining FGR from high burned PWR type fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Miyanishi, Hideyuki; Kitagawa, Isamu; Iida, Shozo; Ito, Tadaharu; Amano, Hidetoshi.

    1991-11-01

    Experimental study was made to evaluate the FGR (Fission Product Gas Release) from high burned PWR type fuel rods by means of non-destructive method through measurement of the gamma activity of 85 Kr isotope which was accumulated in the fuel top plenum. Experimental result shows that it is possible to know the amounts of FGR at fuel plenum by the equations given in the followings. FGR = 0.28C/V f or FGR = 0.07C where, FGR (%) is the amounts of Xe and Kr released from UO 2 fuel, C (counts/h) the radioactivity of 85 Kr at plenum of the tested fuel rod and V f (ml) the plenum volume of the tested fuel rod, respectively. The present study was made by using 14 x 14 PWR type fuel rods preirradiated up to the burn-up of 42.1 MWd/kgU, followed by the pulse irradiation at Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute (JAERI). The FGR of the tested segmented fuel rods were measured by puncturing and found to range from 0.6% to 12% according to the magnitude of the deposited energy given by pulse. Estimated experimental error bands against the above equations were within plus minus 30%. (author)

  13. The influence of weather and fuel type on the fuel composition of the area burned by forest fires in Ontario, 1996-2006.

    Science.gov (United States)

    Podur, Justin J; Martell, David L

    2009-07-01

    Forest fires are influenced by weather, fuels, and topography, but the relative influence of these factors may vary in different forest types. Compositional analysis can be used to assess the relative importance of fuels and weather in the boreal forest. Do forest or wild land fires burn more flammable fuels preferentially or, because most large fires burn in extreme weather conditions, do fires burn fuels in the proportions they are available despite differences in flammability? In the Canadian boreal forest, aspen (Populus tremuloides) has been found to burn in less than the proportion in which it is available. We used the province of Ontario's Provincial Fuels Database and fire records provided by the Ontario Ministry of Natural Resources to compare the fuel composition of area burned by 594 large (>40 ha) fires that occurred in Ontario's boreal forest region, a study area some 430,000 km2 in size, between 1996 and 2006 with the fuel composition of the neighborhoods around the fires. We found that, over the range of fire weather conditions in which large fires burned and in a study area with 8% aspen, fires burn fuels in the proportions that they are available, results which are consistent with the dominance of weather in controlling large fires.

  14. 14 CFR 291.44 - BTS Schedule P-12(a), Fuel Consumption by Type of Service and Entity.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false BTS Schedule P-12(a), Fuel Consumption by... TRANSPORTATION Reporting Rules § 291.44 BTS Schedule P-12(a), Fuel Consumption by Type of Service and Entity. (a.... (e)(1) The cost of fuel shall include shrinkage, but excludes: (i) “Throughput” and “in to plane...

  15. Experience of air transport of nuclear fuel material as type A package

    International Nuclear Information System (INIS)

    Kawasaki, Masashi; Kageyama, Tomio; Suzuki, Toru

    2004-01-01

    Special law on nuclear disaster countermeasures (hereafter called as to nuclear disaster countermeasures low) that is domestic law for dealing with measures for nuclear disaster, was enforced in June, 2000. Therefore, nuclear enterprise was obliged to report accidents as required by nuclear disaster countermeasures law, besides meeting the technical requirement of existent transport regulation. For overseas procurement of plutonium reference materials that are needed for material accountability, A Type package must be transported by air. Therefore, concept of air transport of nuclear fuel materials according to the nuclear disaster countermeasures law was discussed, and the manual including measures against accident in air transport was prepared for the oversea procurement. In this presentation, the concept of air transport of A Type package containing nuclear fuel materials according to the nuclear disaster countermeasures law, and the experience of a transportation of plutonium solution from France are shown. (author)

  16. Cryogenic distillation: a fuel enrichment system for near-term tokamak-type D-T fusion reactors

    International Nuclear Information System (INIS)

    Misra, B.; Davis, J.F.

    1980-02-01

    The successful operation and economic viability of deuterium-tritium- (D-T-) fueled tokamak-type commercial power fusion reactors will depend to a large extent on the development of reliable tritium-containment and fuel-recycle systems. Of the many operating steps in the fuel recycle scheme, separation or enrichment of the isotropic species of hydrogen by cryogenic distillation is one of the most important. A parametric investigation was carried out to study the effects of the various operating conditions and the composition of the spent fuel on the degree of separation. A computer program was developed for the design and analysis of a system of interconnected distillation columns for isotopic separation such that the requirements of near-term D-T-fueled reactors are met. The analytical results show that a distillation cascade consisting of four columns is capable of reprocessing spent fuel varying over a wide range of compositions to yield reinjection-grade fuel with essentially unlimited D/T ratio

  17. 40 CFR 600.209-08 - Calculation of vehicle-specific 5-cycle fuel economy values for a model type.

    Science.gov (United States)

    2010-07-01

    ...-cycle fuel economy values for a model type. 600.209-08 Section 600.209-08 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy Regulations for 1977 and Later Model Year Automobiles-Procedures for...

  18. Operating method of molten carbonate type fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Nakanishi, Tsuneo

    1988-12-06

    Molten carbonate type fuel cell involves a problem of oxidation of anode while the unit is stopped. Although there is a method proposed wherein an inactive gas is supplied to anode during the stoppage, the market-available inactive gas contains a slight amount of oxygen which makes it difficult to prevent the deterioration of the anode. In this invention, at the start and the stop other than the normal operation, a protective gas mixture of an inactive gas with a small amount of hydrogen is supplied to the anode. The inactive gas is a commercial type nitrogen, argon or helium; hydrogen is mixed in amount 0.5 - 2.0% of the inactive gas. By this method, oxygen in air which comes in from the gas-sealed portion of the cell is reduced by hydrogen in the protective gas and is discharged in the form of water. 2 figs.

  19. Numerical study of the thermo-hydraulic behavior for the Candu type fuel channel

    International Nuclear Information System (INIS)

    Lazaro, Pavel Gabriel; Balas Ghizdeanu, Elena Nineta

    2008-01-01

    Candu type reactors use fuel channel in a horizontal lattice. The fuel bundles are positioned in two Zircaloy tubes: the pressure tube surrounded by calandria tube. Inside the pressure tube the coolant heavy water flows. The coolant reaches high temperatures and pressures. Due to irregular neutron spatial distribution, the fuel channel stress differs from one channel to other. In one improbable event of severe accident, the fuel channel behaves differently according to its normal function history. Over the years, there have been many research projects trying to analyze thermal hydraulic performance of the design and to add some operational improvements in order to achieve an efficient thermal hydraulic distribution. This paper discusses the thermo hydraulic behavior (influence of the temperature and velocity distribution) of the most solicited channel, simulated with Fluent 6.X. Code. Moreover it will be commented the results obtained using different models and mesh applied. (authors)

  20. The first critical experiment with a new type of fuel assemblies IRT-3M on the training reactor VR-I

    International Nuclear Information System (INIS)

    Matejka, Karel; Sklenka, Lubomir

    1997-01-01

    The paper 'The first critical experiment with a new type of fuel assemblies IRT-3M on training reactor VR-1 presents basic information about the replacement of fuel on the reactor VR-1 run on FJFI CVUT in Prague. In spring 1997 the IRT-2M fuel type used till then was replaced by the IRT-3M type. When the fuel was replaced, no change in its enrichment was made, i.e. its level remained as 36% 235 U. The replacement itself was carried out in tight co-operation with the Nuclear Research Institute Rez plc., as related to the operation of the research reactor LVR-15. The fuel replacement on the VR-I reactor is a part of the international program RERTR (Reduced Enrichment for Research and Test Reactors) in which the Czech Republic participates. (author)

  1. A summary report on recruitment type researches on nuclear fuel cycle in fiscal year of 2001

    International Nuclear Information System (INIS)

    2002-07-01

    The promotion system on recruitment type researches on nuclear fuel cycle begun on fiscal year of 1999, aims to intend to activate researching environment of the Japan Nuclear Cycle Development Institute (JNC) through intercourses, information exchanges, publication of research results, and so on between researchers in other organizations and JNC, as a result, to effectively promote fundamental and basic R and Ds. This report contains summaries of 28 items of research results on the recruitment type researches on nuclear fuel cycle as 9 items relating to fast breeder reactors, 8 items relating to nuclear fuel cycle, 1 item relating to radiation safety, and 10 items relating to geological disposal and science, carried out on fiscal year of 2001. (G.K.)

  2. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    Energy Technology Data Exchange (ETDEWEB)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro, E-mail: duvan.castellanos@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: pedro.rossi@ufabc.edu.br, E-mail: pedro.carajilescov10@gmail.com [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil). Centro de Engenharias, Modelagem e Ciências Sociais Aplicadas

    2017-07-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  3. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    International Nuclear Information System (INIS)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro

    2017-01-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  4. Investigations on burning efficiency and exhaust emission of in-line type emulsified fuel system

    Energy Technology Data Exchange (ETDEWEB)

    Tseng, Y.K. [National Chinyi University of Technology (Taiwan). Dept. of Mechanical Engineering; Cheng, H.C. [Point Environmental Protection Technology Company Limited (Taiwan)

    2011-07-28

    In this research, the burning efficiency as well as exhaust emission of a new water-in-oil emulsified fuel system was studied. This emulsified system contains two core processes, the first one is to mix 97% water with 3% emulsifier by volume, and get the milk-like emulsified liquid, while the second one is to compound the milk-like emulsified liquid with heavy oil then obtain the emulsified fuel. In order to overcome the used demulsification problem during in reserve or in transport, this system was designed as a made and use in-line type. From the results of a series of burning tests, the fuel saving can be 8--15%. Also, from the comparison of decline for the heat value and total energy output of emulsified fuel, one can find that the water as the dispersed phase in the combustion process will lead to a micro-explosion as well as the water gas effect, both can raise the combustion temperature and burning efficiency. By comparing the waste gas emission of different types of emulsified fuel, one can know that, the CO2 emission reduces approximately 14%, and NOx emission reduces above 46%, meaning the reduction of the exhaust gas is truly effective. From the exhaust temperature of tail pipe, the waste heat discharge also may reduce 27%, it is quite advantageous to the global warming as well as earth environmental protection.

  5. The effects of alternative fuel types on the organoleptic qualities of ...

    African Journals Online (AJOL)

    This study was carried out to assess the effects of alternative smoking fuel types on the organoleptic qualities of coarse pork sausages. The sausages were produced with lean pork (2.5 kg) and pork fat (0.5 kg), minced, mixed with spices and stuffed into natural casings. They were grouped into four and each group was ...

  6. Fuel loading method to exchangeable reactor core of BWR type reactor and its core

    International Nuclear Information System (INIS)

    Koguchi, Kazushige.

    1995-01-01

    In a fuel loading method for an exchangeable reactor core of a BWR type reactor, at least two kinds of fresh fuel assemblies having different reactivities between axial upper and lower portions are preliminarily prepared, and upon taking out fuel assemblies of advanced combustion and loading the fresh fuel assemblies dispersingly, they are disposed so as to attain a predetermined axial power distribution in the reactor. At least two kinds of fresh fuel assemblies have a content of burnable poisons different between the axial upper portion and lower portions. In addition, reactivity characteristics are made different at a region higher than the central boundary and a region lower than the central boundary which is set within a range of about 6/24 to 16/24 from the lower portion of the fuel effective length. There can be attained axial power distribution as desired such as easy optimization of the axial power distribution, high flexibility, and flexible flattening of the power distribution, and it requires no special change in view of the design and has a good economical property. (N.H.)

  7. Dissolution process for advanced-PWR-type fuels

    International Nuclear Information System (INIS)

    Black, D.E.; Decker, L.A.; Pearson, L.G.

    1979-01-01

    The new Fluorinel Dissolution Process and Fuel Storage (FAST) Facility at ICPP will provide underwater storage of spent PWR fuel and a new head-end process for fuel dissolution. The dissolution will be two-stage, using HF and HNO 3 , with an intermittent H 2 SO 4 dissolution for removing stainless steel components. Equipment operation is described

  8. Improvement of visualization efficiency for the nondestructive inspection image of internal defects in plate type nuclear fuel

    International Nuclear Information System (INIS)

    Park, Seung Kyu; Park, Nak Kyu; Baik, Sung Hoon; Lee, Yoon Sang; Cheong, Yong Moo; Kang, Young June

    2012-01-01

    Plate type nuclear fuel has been adopted in most research reactors. The production quality of the fuel is a key part for an efficient and stable generation of thermal energy in research reactors. Thus, a nondestructive quality inspection for the internal defects of plate type nuclear fuel is a key process during the production of nuclear fuel for safety insurance. Nondestructive quality inspections based on X rays and ultrasounds have been widely used for the defect detection of plate type nuclear fuel. X ray testing is a simple and fast inspection method, and provides an image in real time as the inspection results. Thus, the testing can be carried out by a non expert field worker. However, it is hard to detect closed type defects that should be detected during the production of plate type nuclear fuel. Ultrasonic testing is a powerful tool to detect internal defects including open type and closed type defects in plate type nuclear fuel. However, the inspection process is complicated because an immersion test should be carried out in a water tank. It is also a time consuming inspection method because area testing to acquire image is based on the scanning of the point by point inspections. Among nondestructive inspection techniques, the techniques based on laser interferometry and infrared thermography have been widely used in the detection of internal defects of plate type composite materials, such as aircraft, automotive etc. While infrared thermography technique (IRT) analyses the thermal behavior of the specimen surface, laser interferometry technique (LIT) analyses the deformation field. Both techniques are useful tools for detection and evaluation of internal defects in composite materials. Especially, the laser interferometry technique can provide the depth information of internal defects. Laser interferometry technique (LIT) is a non contact inspection method faster than thermography. Also, this technique requires less energy than thermography and the

  9. In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

    International Nuclear Information System (INIS)

    Gunnerson, F.S.; MacDonald, P.E.

    1980-01-01

    The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made

  10. Development of a Direct Methanol Fuel Cell with Lightweight Disc Type Current Collectors

    Directory of Open Access Journals (Sweden)

    Yean-Der Kuan

    2014-05-01

    Full Text Available The direct methanol fuel cell (DMFC adopts methanol solution as a fuel suitable for low power portable applications. A miniature, lightweight, passive air-breathing design is therefore desired. This paper presents a novel planar disc-type DMFC with multiple cells containing a novel developed lightweight current collector at both the anode and cathode sides. The present lightweight current collector adopts FR4 Glass/Epoxy as the substrate with the current collecting areas located at the corresponding membrane electrolyte assembly (MEA areas. The current collecting areas are fabricated by sequentially coating a corrosion resistant layer and electrical conduction layer via the thermal evaporation technique. The anode current collector has carved flow channels for fuel transport and production. The cathode current collector has drilled holes for passive air breathing. In order to ensure feasibility in the present concept a 3-cell prototype DMFC module with lightweight disc type current collectors is designed and constructed. Experiments were conducted to measure the cell performance. The results show that the highest cell power output is 54.88 mW·cm−2 and successfully demonstrate the feasibility of this novel design.

  11. Thermal characteristics during hydrogen fueling process of type IV cylinder

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Chan [Department of Fire and Disaster Prevention, Kyungil University, 33, Buhori, Hayang, Kyungsan 712-701 (Korea); Lee, Seung Hoon; Yoon, Kee Bong [Department of Mechanical Engineering, Chung Ang University, 221, Huksuk, Dongjak, Seoul 156-756 (Korea)

    2010-07-15

    Temperature increase during hydrogen fueling process is a significant safety concern of a high pressure hydrogen vessel. Hence, thermal characteristics of a Type IV cylinder during hydrogen filling process need to be understood. In this study, a series of experiments were conducted to quantify the temperature change of the cylinder during hydrogen filling to 35 MPa. Computational fluid dynamics (CFD) analysis was also conducted to simulate the conditions of the experiments. The results predicted by the CFD analysis show reasonable agreement with the experiments and the discrepancy between the CFD results and experimental results decrease with higher initial gas pressures. The upper and the lower parts of the vessel showed a temperature difference in the vertical direction. The upper gas temperature was higher than that of the lower part due to the buoyancy effect in the vessel. The maximum gas temperature was higher than the maximum temperature allowed in the ISO safety code (85 C) for the case in which the vessel was pressurized from 0 MPa to 35 MPa. This work contributes to the understanding of the thermal flow characteristics of the hydrogen filling process and notes that additional efforts should be made to guarantee the safety of a type IV cylinder during the hydrogen fueling process. (author)

  12. Fuel rod for use in BWR type reactor

    International Nuclear Information System (INIS)

    Takeuchi, Kiyoshi.

    1989-01-01

    A hollow intermediate end plug is disposed to a plenum portion of a fuel rod and a plenum spring is disposed between the end plug and the upper end of a fuel pellet. Then, a hollow portion is disposed between the intermediate end plug and an upper end plug. Thus, since a only a non exothermic portion is present from the intermediate end plug to the upper end plug, oxidation, corrosion, etc. to the fuel can are not caused so much as in the exothermic portion. Accordingly, the wall thickness of the fuel may be reduced to such a extent as only capable of withstanding the external pressure by coolants and the increasing inner pressure due to the release of FP gases and, accordingly, the wall thickness can be reduced as compared with that of the fuel portion in the fuel can. Further, since the power density per unit length of the fuel rod is reduced for fuels with increased number of fuel rods, it is possible to design so as to reduce the release amount of FP gases thereby decreasing the plenum volume. Further, since the surface area in the coolant phase stream portion is reduced, it can be expected for decreasing the pressure loss of fuels and accompanying effect for improving the channel stability. (T.M.)

  13. Guia prático de excercícios de alongamento como promotor de autonômia em um grupo de mulheres / Promoting autonomy in a group of women with a practical guide to stretching

    Directory of Open Access Journals (Sweden)

    Tialhes Farias Marconato

    2018-04-01

    Full Text Available A Terapia Ocupacional desenvolve sua prática e estudo sobre as atividades humanas com a utilização de recursos e instrumentos terapêuticos, que podem beneficiar o processo de envelhecimento. Nesse sentido, o estudo visou reconhecer e descrever a percepção de mulheres com idade entre 40 e 60 anos de idade, participantes de um grupo de Terapia Ocupacional, quanto às repercussões da utilização de um Guia Prático de Alongamento, no desempenho das atividades cotidianas. Para tanto, realizou-se uma pesquisa qualitativa, exploratória e descritiva, por meio de Grupo Focal. Verificou-se que a utilização do Guia influenciou no controle da dor, na autonomia para realização de atividades de vida diária e na percepção do próprio corpo das participantes do estudo. Conclui-se que o Guia Prático de Alongamento é uma ferramenta simples, porém relevante para melhorar a consciência corporal, a autonomia e independência de indivíduos em processo de envelhecimento, sugerindo-o como um instrumento indicado para outras populações. AbstractOccupational Therapy carries out the practice and study of human activities by employing therapeutic resources and instruments, which can benefit the aging process. Thus, the aim of this study was to identify and describe the perception of women between the ages of 40 and 60 in an Occupational Therapy group regarding the effects of using a Practical Guide to Stretching on the performance of their daily activities. In that sense, a qualitative, exploratory and descriptive Focus Group research was carried out. It was verified that the use of the Practical Guide had an influence on participants’ pain control, autonomy to perform daily activities and body perception. It was concluded that the Practical Guide to Stretching is a simple but relevant tool that helps aging individuals improve their body awareness, autonomy and independence. It is also a suitable tool for other populations.Keywords: Muscle

  14. Fuel performance of DOE fuels in water storage

    International Nuclear Information System (INIS)

    Hoskins, A.P.; Scott, J.G.; Shelton-Davis, C.V.; McDannel, G.E.

    1993-01-01

    Westinghouse Idaho Nuclear Company operates the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory. In April of 1992, the U.S. Department of Energy (DOE) decided to end the fuel reprocessing mission at ICPP. Fuel performance in storage received increased emphasis as the fuel now needs to be stored until final dispositioning is defined and implemented. Fuels are stored in four main areas: an original underwater storage facility, a modern underwater storage facility, and two dry fuel storage facilities. As a result of the reactor research mission of the DOE and predecessor agencies, the Energy Research and Development Administration and the Atomic Energy Commission, many types of nuclear fuel have been developed, used, and assigned to storage at the ICPP. Fuel clad with stainless steel, zirconium, aluminum, and graphite are represented. Fuel matrices include uranium oxide, hydride, carbide, metal, and alloy fuels, resulting in 55 different fuel types in storage. Also included in the fuel storage inventory is canned scrap material

  15. The fuel to clad heat transfer coefficient in advanced MX-type fuel pins

    International Nuclear Information System (INIS)

    Caligara, F.; Campana, M.; Mandler, R.; Blank, H.

    1979-01-01

    Advanced fuels (mixed carbides, nitrides and carbonitrides) are characterised by a high thermal conductivity compared to that of oxide fuels (5 times greater) and their behaviour under irradiation (amount of swelling, fracture behaviour, restructuring) is far more sensitive to the design parameters and to the operating temperature than that of oxide fuels. The use of advanced fuels is therefore conditioned by the possibility of mastering the above phenomena, and the full exploitation of their favorable neutron characteristics depends upon a good understanding of the mutual relationships of the various parameters, which eventually affect the mechanical stability of the pin. By far the most important parameter is the radial temperature profile which controls the swelling of the fuel and the build-up of stress fields within the pin. Since the rate of fission gas swelling of these fuels is relatively large, a sufficient amount of free space has to be provided within the pin. This space originally appears as fabrication porosity and as fuel-to-clad clearance. Due to the large initial gap width and to the high fuel thermal conductivity, the range of the fuel operating temperatures is mainly determined by the fuel-to-clad heat transfer coefficient h, whose correct determination becomes one of the central points in modelling. During the many years of modelling activity in the field of oxide fuels, several theoretical models have been developed to calculate h, and a large amount of experimental data has been produced for the empirical adjustment of the parameters involved, so that the situation may be regarded as rather satisfactory. The analysis lead to the following conclusions. A quantitative comparison of experimental h-values with existing models for h requires rather sophisticated instrumented irradiation capsules, which permit the measurement of mechanical data (concerning fuel and clad) together with heat rating and temperatures. More and better well

  16. Fuel assembly for use in BWR type reactor

    International Nuclear Information System (INIS)

    Inaba, Yuzo.

    1988-01-01

    Purpose: To attain the reduction of neutron irradiation amount to control rods by the improvement in the reactor shutdown margin and the improvement of the control rod worth, by enhancing the arrangement of burnable poisons. Constitution: The number of burnable poison-incorporated fuel rods present in the outer two rows along the sides in adjacent with a control rod among the square lattice arrangement in a fuel assembly is decreased to less than 1/4 for that of total burnable poison-incorporated fuel rods, while the remaining burnable posion-incorporated fuel rods are arranged in the region other than above (that is, those regions not nearer to the control rod). Thus, even if a sufficient number of burnable poison to prolong the controlling effect for the reactivity with the burnable contents as the fuel assembly are disposed, only the burnable poison -incorporated fuel rods by the number less than 1/4 for that of the total burnable poison-incorporated fuel rods are present near the control rod of the fuel assembly. Accordingly, the control rod worth at the initial stage of the burning is increased at both high and normal temperatures. (Kawakami, Y.)

  17. Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.; Billone, M.C.; Holland, J.W.; Kramer, J.M.

    1993-01-01

    Three furnace heating tests were conducted with irradiated, HT9-clad and U-19wt%Pu-10wt%Zr-alloy, EBR-II Mk-V-type fuel elements to evaluate the behavior that could be expected during a loss-of-flow event in the reactor. In general, very significant safety margins for cladding breaching have been demonstrated in these tests, under conditions that would envelop a bounding unlikely loss-of-flow event in EBR-II. Highlights of the test results are presented, as are discussions of the cladding breaching mechanisms, axial fuel motion, and fuel surface liquefaction that were found in these tests. (orig.)

  18. Fuel consumption models for pine flatwoods fuel types in the southeastern United States

    Science.gov (United States)

    Clinton S. Wright

    2013-01-01

    Modeling fire effects, including terrestrial and atmospheric carbon fluxes and pollutant emissions during wildland fires, requires accurate predictions of fuel consumption. Empirical models were developed for predicting fuel consumption from fuel and environmental measurements on a series of operational prescribed fires in pine flatwoods ecosystems in the southeastern...

  19. Method of fabricating zirconium metal for use in composite type fuel cans

    International Nuclear Information System (INIS)

    Imahashi, Hiromichi; Inagaki, Masatoshi; Akabori, Kimihiko; Tada, Naofumi; Yasuda, Tetsuro.

    1985-01-01

    Purpose: To mass produce zirconium metal for fuel cans with less radiation hardening. Method: Zirconium sponges as raw material are inserted in a hearth mold and a procedure of melting the zirconium sponges portionwise by using a melting furnace having electron beams as a heat source while moving the hearth is repeated at least for once. Then, the rod-like ingot after melting is melted again in a vacuum or inert gas atmosphere into an ingot of a low oxygen density capable of fabrication. A composite fuel can billet is formed by using the thus obtained zirconium ingot and a zircalloy, and a predetermined composite type fuel can is manufactured by way of hot extrusion and pipe drawing fabrication. The raw material usable herein is zirconium sponge with an oxygen density of 400 ppm or higher and the content of impurity other than oxygen is between 1000 - 5000 ppm in total, or the molten material thereof. (Kamimura, M.)

  20. Constant strength fuel-fuel cell

    International Nuclear Information System (INIS)

    Vaseen, V.A.

    1980-01-01

    A fuel cell is an electrochemical apparatus composed of both a nonconsumable anode and cathode; and electrolyte, fuel oxidant and controls. This invention guarantees the constant transfer of hydrogen atoms and their respective electrons, thus a constant flow of power by submergence of the negative electrode in a constant strength hydrogen furnishing fuel; when said fuel is an aqueous absorbed hydrocarbon, such as and similar to ethanol or methnol. The objective is accomplished by recirculation of the liquid fuel, as depleted in the cell through specific type membranes which pass water molecules and reject the fuel molecules; thus concentrating them for recycle use

  1. PLACA/DPLACA: a code to simulate the behavior of a monolithic/dispersed plate type fuel

    International Nuclear Information System (INIS)

    Denis, Alicia; Soba, Alejandro

    2005-01-01

    The PLACA code was originally built to simulate monolithic plate fuels contained in a metallic cladding, with a gap in between. The international program of high density fuels was recently oriented to the development of a plate-type fuel of a uranium rich alloy with a molybdenum content between 6 to 10 w %, without gap and with a Zircaloy cladding. To give account of these fuels, the DPLACA code was elaborated as a modification of the original code. The extension of the calculation tool to disperse fuels involves a detailed study of the properties and models (still in progress). Of special interest is the material formed by U Mo particles dispersed in an Al matrix. This material has appeared as a candidate fuel for high flux research reactors. However, the interaction layer that grows around the particles has a deleterious effect on the material performance in operation conditions and may represent a limit for its applicability. A number of recent experiments carried out on this material provide abundant information that allows testing of the numerical models. (author)

  2. The theoretical possibility of reducing the doubling time in a fast-reactor by using heterogeneous configurations of various types of fuel

    International Nuclear Information System (INIS)

    Orlov, V.V.; Slesarev, I.S.; Zaritskij, S.M.; Subbotin, S.A.; Alekseev, P.N.; Zverkov, Yu.A.

    1980-01-01

    The authors have derived approximate expressions relating the doubling time of a fast reactor using various types of fuel simultaneously to the doubling time of traditional (homogeneous) reactors in which these types of fuel are used separately. These relationships afford a means of determining the conditions in which the use of various types of fuel can result in an improved doubling time. It was established that the use of heterogeneous compositions formed from assemblies of homogeneous systems gives a notable gain in doubling time over that of any of the original homogeneous systems if the doubling times were similar to each other. This gain is fairly large even in the case of BN reactors with high fuel volume fractions. The size of the gain depends on the degree of ''differentiation'' in the neutron and thermal properties of the components of the heterogeneous reactor. An optimum proportion has been found for the assemblies taken from the original homogeneous systems, governed primarily by the ratio of fuel densities. Estimates were made of the advantages of metallic oxide compositions over the traditional compositions used in large, fast reactors of the BN type. These estimates indicate that the former can be considered as alternative homogeneous compositions with carbide or nitride fuel as far as breeding characteristics are concerned. (author)

  3. Tensile Test of Welding Joint Parts for a Plate-type Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Kim, J. Y.; Kim, H. J.; Yim, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The tensile tests were performed using an INSTRON 4505 (universal tensile) testing machine. These welding joints are composed of two parts for the soundness of the fuel assembly; one is the side plate with a fixing bar and the other is a side plate with an end fitting. These two joint parts are fabricated by TIG welding method. The tensile tests of the welding joints of a plate-type FA are executed by a tensile test. The fixture configurations for the specimen are very important to obtain the strict test results. The maximum strength has an approximately linear correlation with the unit bonding length of the welding joints. In spite of these results, the maximum strengths of the welding joints are satisfied according to the minimum requirement. These tensile tests of the joint parts for a plate-type fuel assembly (FA) have to be executed to evaluate the structural strength. For the tensile test, the joint parts of a FA used in the test are made of aluminum alloy (Al6061-T6)

  4. Tensile Test of Welding Joint Parts for a Plate-type Fuel Assembly

    International Nuclear Information System (INIS)

    Yoon, K. H.; Kim, J. Y.; Kim, H. J.; Yim, J. S.

    2013-01-01

    The tensile tests were performed using an INSTRON 4505 (universal tensile) testing machine. These welding joints are composed of two parts for the soundness of the fuel assembly; one is the side plate with a fixing bar and the other is a side plate with an end fitting. These two joint parts are fabricated by TIG welding method. The tensile tests of the welding joints of a plate-type FA are executed by a tensile test. The fixture configurations for the specimen are very important to obtain the strict test results. The maximum strength has an approximately linear correlation with the unit bonding length of the welding joints. In spite of these results, the maximum strengths of the welding joints are satisfied according to the minimum requirement. These tensile tests of the joint parts for a plate-type fuel assembly (FA) have to be executed to evaluate the structural strength. For the tensile test, the joint parts of a FA used in the test are made of aluminum alloy (Al6061-T6)

  5. Calculational modeling of fuel assemblies of WWER-1000 type with the use of burnable absorber Gadolinum; comparative analysis

    International Nuclear Information System (INIS)

    Yeremenko, M.L.; Kovbasenko, Yu.P.; Loetsch, T.

    2001-01-01

    In connection with the beginning of the use of fuel assemblies with burnable absorbers by integration of Gadolinum into the nuclear fuel at Ukrainian NPP the task of testing the code systems and the pertinent neutron cross section libraries for the new fuel arose. Taking into account the long term experience of German experts with calculations and evaluation of nuclear fuel containing Gadolinum it was decided to carry out a series of test calculations for fuel assembly lattices of PWR, WWER-440 and WWER-1000 types using the NESSEL/PYTHIA and CASMO/SIMULATE code systems (Authors)

  6. Effects of spent fuel types on offsite consequences of hypothetical accidents

    International Nuclear Information System (INIS)

    Courtney, J. C.; Dwight, C. C.; Lehto, M. A.

    2000-01-01

    Argonne National Laboratory (ANL) conducts experimental work on the development of waste forms suitable for several types of spent fuel at its facility on the Idaho National Engineering and Environmental Laboratory (INEEL) located 48 km West of Idaho Falls, ID. The objective of this paper is to compare the offsite radiological consequences of hypothetical accidents involving the various types of spent nuclear fuel handled in nonreactor nuclear facilities. The highest offsite total effective dose equivalents (TEDEs) are estimated at a receptor located about 5 km SSE of ANL facilities. Criticality safety considerations limit the amount of enriched uranium and plutonium that could be at risk in any given scenario. Heat generated by decay of fission products and actinides does not limit the masses of spent fuel within any given operation because the minimum time elapsed since fissions occurred in any form is at least five years. At cooling times of this magnitude, fewer than ten radionuclides account for 99% of the projected TEDE at offsite receptors for any credible accident. Elimination of all but the most important nuclides allows rapid assessments of offsite doses with little loss of accuracy. Since the ARF (airborne release fraction), RF (respirable fraction), LPF (leak path fraction) and atmospheric dilution factor (χ/Q) can vary by orders of magnitude, it is not productive to consider nuclides that contribute less than a few percent of the total dose. Therefore, only 134 Cs, 137 Cs- 137m Ba, and the actinides significantly influence the offsite radiological consequences of severe accidents. Even using highly conservative assumptions in estimating radiological consequences, they remain well below current Department of Energy guidelines for highly unlikely accidents

  7. Installation, maintenance and operating manual for the Lucas-type fuel injection system of the 3 B rotary engine

    Science.gov (United States)

    1985-01-01

    The installation procedure, maintenance, adjustment and operation of a Lucas type fuel injection system for 13B rotary racing engine is outlined. Components of the fuel injection system and installation procedure and notes are described. Maintenance, adjustment, and operation are discussed.

  8. Analysis of fuel management pattern of research reactor core of the MTR type design

    International Nuclear Information System (INIS)

    Lily Suparlina; Tukiran Surbakti

    2014-01-01

    Research reactor core design needs neutronics parameter calculation use computer codes. Research reactor MTR type is very interested because can be used as research and also a radioisotope production. The research reactor in Indonesia right now is already 25 years old. Therefore, it is needed to design a new research reactor as a compact core. Recent research reactor core is not enough to meet criteria acceptance in the UCD which already determined namely thermal neutron flux in the core is 1.0x10 15 n/cm 2 s. so that it is necessary to be redesign the alternative core design. The new research reactor design is a MTR type with 5x5 configuration core, uses U9Mo-Al fuel, 70 cm of high and uses two certainly fuel management pattern. The aim of this research is to achieve neutron flux in the core to meet the criteria acceptance in the UCD. Calculation is done by using WIMSD-B, Batan-FUEL and Batan-3DIFF codes. The neutronic parameters to be achieved by this calculation are the power level of 50 MW thermal and core cycle of 20 days. The neutronics parameter calculation is done for new U-9Mo-Al fuel with variation of densities.The result of calculation showed that the fresh core with 5x5 configuration, 360 gram, 390 gram and 450 gram of fuel loadings have meet safety margin and acceptance criteria in the UCD at the thermal neutron flux is more then 1.0 x 10 15 n/cm 2 s. But for equilibrium core is only the 450 gram of loading meet the acceptance criteria. (author)

  9. Effect of in-pile degradation of the meat thermal conductivity on the maximum temperature of the plate-type U-Mo dispersion fuels

    International Nuclear Information System (INIS)

    Medvedev, Pavel G.

    2009-01-01

    Effect of in-pile degradation of thermal conductivity on the maximum temperature of the plate-type research reactor fuels has been assessed using the steady-state heat conduction equation and assuming convection cooling. It was found that due to very low meat thickness, characteristic for this type of fuel, the effect of thermal conductivity degradation on the maximum fuel temperature is minor. For example, the fuel plate featuring 0.635 mm thick meat operating at heat flux of 600 W/cm2 would experience only a 20 C temperature rise if the meat thermal conductivity degrades from 0.8 W/cm-s to 0.3 W/cm-s. While degradation of meat thermal conductivity in dispersion-type U-Mo fuel can be very substantial due to formation of interaction layer between the particles and the matrix, and development of fission gas filled porosity, this simple analysis demonstrates that this phenomenon is unlikely to significantly affect the temperature-based safety margin of the fuel during normal operation.

  10. Influence of the voids fraction in the power distribution for two different types of fuel assemblies

    International Nuclear Information System (INIS)

    Jacinto C, S.; Del Valle G, E.; Alonso V, G.; Martinez C, E.

    2017-09-01

    In this work an analysis of the influence of the voids fraction in the power distribution was carried out, in order to understand more about the fission process and the energy produced by the fuel assembly type BWR. The fast neutron flux was analyzed considering neutrons with energies between 0.625 eV and 10 MeV. Subsequently, the thermal neutron flux analysis was carried out in a range between 0.005 eV and 0.625 eV. Likewise, its possible implications in the power distribution of the fuel cell were also analyzed. These analyzes were carried out for different void fraction values: 0.2, 0.4 and 0.8. The variations in different burn steps were also studied: 20, 40 and 60 Mwd / kg. These values were studied in two different types of fuel cells: Ge-12 and SVEA-96, with an average initial enrichment of 4.11%. (Author)

  11. Methodology for thermal-hydraulics analysis of pool type MTR fuel research reactors

    International Nuclear Information System (INIS)

    Umbehaun, Pedro Ernesto

    2000-01-01

    This work presents a methodology developed for thermal-hydraulic analysis of pool type MTR fuel research reactors. For this methodology a computational program, FLOW, and a model, MTRCR-IEAR1 were developed. FLOW calculates the cooling flow distribution in the fuel elements, control elements, irradiators, and through the channels formed among the fuel elements and among the irradiators and reflectors. This computer program was validated against experimental data for the IEA-R1 research reactor core at IPEN-CNEN/SP. MTRCR-IEAR1 is a model based on the commercial program Engineering Equation Solver (EES). Besides the thermal-hydraulic analyses of the core in steady state accomplished by traditional computational programs like COBRA-3C/RERTR and PARET, this model allows to analyze parallel channels with different cooling flow and/or geometry. Uncertainty factors of the variables from neutronic and thermalhydraulic calculation and also from the fabrication of the fuel element are introduced in the model. For steady state analyses MTRCR-IEAR1 showed good agreement with the results of COBRA-3C/RERTR and PARET. The developed methodology was used for the calculation of the cooling flow distribution and the thermal-hydraulic analysis of a typical configuration of the IEA-R1 research reactor core. (author)

  12. The use of U3Si2 dispersed in aluminum in plate-type fuel elements for research and test reactors

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Domagala, R.F.; Hofman, G.L.; Wiencek, T.C.; Copeland, G.L.; Hobbs, R.W.; Senn, R.L.

    1987-10-01

    A high-density fuel based on U 3 Si 2 dispersed in aluminum has been developed and tested for use in converting plate-type research and test reactors from the use of highly enriched uranium to the use of low-enriched uranium. Results of preirradiation testing and the irradiation and postirradiation examination of miniature fuel plates and full-sized fuel elements are summarized. Swelling of the U 3 Si 2 fuel particles is a linear function of the fission density in the particle to well beyond the fission density achievable in low-enriched fuels. U 3 Si 2 particle swelling rate is approximately the same as that of the commonly used UAl/sub x/ fuel particle. The presence of minor amounts of U 3 Si or uranium solid solution in the fuel result in greater, but still acceptable, fuel swelling. Blister threshold temperatures are at least as high as those of currently used fuels. An exothermic reaction occurs near the aluminum melting temperature, but the measured energy releases were low enough not to substantially worsen the consequences of an accident. U 3 Si 2 -aluminum dispersion fuel with uranium densities up to at least 4.8 Mg/m 3 is a suitable LEU fuel for typical plate-type research and test reactors. 42 refs., 28 figs., 7 tabs

  13. Evaluation of plate type fuel options for small power reactors; Avaliacao de alternativas de combustivel tipo placa para reatores de pequeno porte

    Energy Technology Data Exchange (ETDEWEB)

    Andrzejewski, Claudio de Sa

    2005-07-01

    Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO{sub 2} in stainless steel, of UO{sub 2} in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

  14. Fuel cells for electricity generation from carbonaceous fuels

    Energy Technology Data Exchange (ETDEWEB)

    Ledjeff-Hey, K; Formanski, V; Roes, J [Gerhard-Mercator- Universitaet - Gesamthochschule Duisburg, Fachbereich Maschinenbau/Fachgebiet Energietechnik, Duisburg (Germany); Heinzel, A [Fraunhofer Inst. for Solar Energy Systems (ISE), Freiburg (Germany)

    1998-09-01

    Fuel cells, which are electrochemical systems converting chemical energy directly into electrical energy with water and heat as by-products, are of interest as a means of generating electricity which is environmentally friendly, clean and highly efficient. They are classified according to the electrolyte used. The main types of cell in order of operating temperature are described. These are: alkaline fuel cells, the polymer electrolyte membrane fuel cell (PEMFC); the phosphoric acid fuel cell (PAFC); the molten carbonate fuel cell (MCFC); the solid oxide fuel cell (SOFC). Applications depend on the type of cell and may range from power generation on a large scale to mobile application in cars or portable systems. One of the most promising options is the PEM-fuel cell stack where there has been significant improvement in power density in recent years. The production from carbonaceous fuels and purification of the cell fuel, hydrogen, is considered. Of the purification methods available, hydrogen separation by means of palladium alloy membranes seems particular effective in reducing CO concentrations to the low levels required for PEM cells. (UK)

  15. Analytical Evaluation to Determine Selected PAHs in a Contaminated Soil With Type II Fuel

    International Nuclear Information System (INIS)

    Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Garcia Frutos, F. J.

    2010-01-01

    A study on the optimization of an ultrasonic extraction method for selected PAHs determination in soil contaminated by type II fuel and by using HPLC with fluorescence detector is presented. The main objective was optimize the analytical procedure, minimizing the volume of solvent and analysis time and avoiding possible loss by evaporation. This work was carried out as part of a project that investigated a remediation process of agricultural land affected by an accidental spillage of fuel (Plan Nacional I + D + i, CTM2007-64 537). The paper is structured as: Optimization of wavelengths in the chromatographic conditions to improve resolution in the analysis of fuel samples. Optimization of the main parameters affecting in the extraction process by sonication. Comparison of results with those obtained by accelerated solvent extraction. (Author) 3 refs.

  16. New Concept of Designing Composite Fuel for Fast Reactors with Closing Fuel Cycle

    International Nuclear Information System (INIS)

    Savchenko, A.; Vatulin, A.; Uferov, O.; Kulakov, G.; Sorokin, V.

    2013-01-01

    For fast reactors a novel type of promising composite U-PuO2 fuel is proposed which is based on dispersion fuel elements. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. Novel fuel features higher characteristics in comparison to metallic or MOX fuel its fabrication technology is readily accomplished and is environmentally clean. A possibility is demonstrated of fabricating coated steel claddings to protect from interaction with fuel and fission products when use standard rod type MOX or metallic U-Pu-Zr fuel. Novel approach to reprocessing of composite fuel is demonstrated, which allows to separate uranium from burnt plutonium as well as the newly generated fissile plutonium from burnt one without chemical processes, which simplifies the closing of the nuclear fuel cycle. Novel composite fuel combines the advantages of metallic and ceramic types of fuel and has high uranium density that allows also to implicate it in BREST types reactor with conversion ratio more than 1. Peculiarities of closing nuclear cycle with composite fuel are demonstrated that allows more effective re-usage of generated Pu as well as, minimizing r/a wastes by incineration of MA in specially developed IMF design

  17. Physics concept on the constellation type fissile fuels and its application to the prospective Th-232U Reactor

    International Nuclear Information System (INIS)

    Zhang, Jiahua

    1994-01-01

    In contrast with the conventional nuclear reactor which usually fuelled with on single fissile nuclide, a constellation type fissile fuels reactor consists of a parent nuclide such as 232 Th or 238 U and its whole family of neutron generated daughter nuclides. All of them are regarded as fissile fuels but of quite different fission ability. The concentration of each daughter nuclide is determined by its saturate concentration ratio with the parent nuclide. In such fuel system, the whole fuel consumed by neutron reaction almost completely results in fission products. In this article, some properties of such fuel system, determination of the saturate concentration of each daughter nuclide and applicability to Th- 233 U fueled reactor will be discussed. 3 refs., 1 tab., 2 figs

  18. Qualitative and quantitative characteristics of fission products in spent nuclear fuel from RBMK-type reactor

    International Nuclear Information System (INIS)

    Adlys, G.; Adliene, D.

    2002-01-01

    Well-known empirical models or experimental instruments and methods for the estimation of fission product yields do not allow prediction of the behavior and evaluation of the time-dependent qualitative and quantitative characteristics of all fission products in spent nuclear fuel during long-term storage. Several computer codes were developed in different countries to solve this problem. French codes APOLLO1 and PEPIN were used in this work for modeling the characteristics of spent nuclear fuel in the RBMK reactor. The modeling results of qualitative and quantitative characteristics of long-lived fission products for different cooling periods of spent nuclear fuel, including 50-year cooling period, are presented in this paper. The 50-year cooling period conforms to the foreseen time of storage of spent nuclear fuel in CONSTOR and CASTOR casks at the Ignalina NPP. These results correlate well with evaluated quantities for the well-known yields of the nuclides and could be used for the compilation of the database for long-lived fission products in spent nuclear fuel from the RBMK-type reactor. They allow one to predict and to solve effectively safety problems concerning with long-term spent nuclear fuel storage in casks. (author)

  19. Heat conduction in a plate-type fuel element with time-dependent boundary conditions

    International Nuclear Information System (INIS)

    Faya, A.J.G.; Maiorino, J.R.

    1981-01-01

    A method for the solution of boundary-value problems with variable boundary conditions is applied to solve a heat conduction problem in a plate-type fuel element with time dependent film coefficient. The numerical results show the feasibility of the method in the solution of this class of problems. (Author) [pt

  20. Fuel assemblies for FBR type reactor

    International Nuclear Information System (INIS)

    Ikeda, Kiyoshi.

    1981-01-01

    Purpose: To decrease errors in the flow rate distribution of coolants by resiliently inserting a flow regulation rod having a variable flow regulation element formed at the upper portion along the axial direction in the entrance nozzle of a fuel assembly. Constitution: A plurality of orifice aperture are formed to the entrance nozzle of a fuel assembly and an aperture for inserting a flow regulation rod is formed to the top end of the entrance nozzle. A fixed flow regulation element A and a variable flow regulation element B supported coaxially with the nozzle by a support ring are disposed to the inside of the nozzle. The element B is urged by the resilient urging spring to the element A and connected by way of support lever to the flow regulation rod. While on the other hand, the top end of the nozzle is inserted through the partition wall between a high pressure coolant chamber and a low pressure coolant chamber. An aperture for hydrodynamically supporting the fuel assembly is provided by way of a frame and a flow regulation rod that stands vertically from the low pressure coolant chamber is disposed to the center of the frame. In the fuel assembly, the flow regulation rod inserted from the aperture at the top end of the nozzle pushes the element B upwardly to thereby maintain a flow passage of the coolant between the elements A and B. (Seki, T.)

  1. Method of determining the composition of fuels for FBR type reactors

    International Nuclear Information System (INIS)

    Tsutsumi, Kiyoshi.

    1981-01-01

    Purpose: To improve the core safety of FBR type reactors by determining the composition of fuels composed of oxide mixture of plutonium and uranium, using a relation between specific plutonium seed and plutonium enrichment degree. Method: Relation is determined between the ratio of a specific plutonium seed for constituting plutonium oxide, for example 239 U ratio and a plutonium enrichment degree required for setting the assembly power to a constant level. The ratio of 239 U is plutonium having a given isotopic ratio is also determined. The accuracy of the 239 U ratio can be improved by the correction using the density coefficient. Then, the plutonium enrichment degree is determined using the relation determined as above based on the thus determined 239 U ratio. The composition of the fuel using oxide mixture of plutonium and uranium is determined by utilizing the thus obtained plutonium enrichment degree. (Moriyama, K.)

  2. Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

    International Nuclear Information System (INIS)

    Roake, W.E.; Adamson, M.G.; Hilbert, R.F.; Langer, S.

    1977-01-01

    Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to ∼60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

  3. Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States); Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States); Hilbert, R F; Langer, S

    1977-04-01

    Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to {approx}60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

  4. Fuel spacer

    International Nuclear Information System (INIS)

    Nishida, Koji; Yokomizo, Osamu; Kanazawa, Toru; Kashiwai, Shin-ichi; Orii, Akihito.

    1992-01-01

    The present invention concerns a fuel spacer for a fuel assembly of a BWR type reactor and a PTR type reactor. Springs each having a vane are disposed on the side surface of a circular cell which supports a fuel rods. A vortex streams having a vertical component are formed by the vanes in the flowing direction of a flowing channel between adjacent cylindrical cells. Liquid droplets carried by streams are deposited on liquid membrane streams flowing along the fuel rod at the downstream of the spacer by the vortex streams. In view of the above, the liquid droplets can be deposited to the fuel rod without increasing the amount of metal of the spacer. Accordingly, the thermal margin of the fuel assembly can be improved without losing neutron economy. (I.N.)

  5. Chemometrics application in fuel's MTR type chemical characterization by X-ray fluorescence

    International Nuclear Information System (INIS)

    Silva, Clayton Pereira da

    2012-01-01

    In Brazil and worldwide the nuclear power has occupied a prominent position with many applications in industry, power generation, environment and medicine, improving the quality of tests and treatments, therefore people's lives. Uranium is the main element used in nuclear facilities and it s employed as base material to generation of electricity in the manufacture of radiopharmaceuticals. In the '50s, during the Cold War, the then newly created International Atomic Energy Agency proposed to oversee nuclear facilities and encourage the manufacture of nuclear fuels with low-enriched uranium (LEU) fuel came then type Material Test Reactor (MTR), manufactured initially in U 3 O 8 and U 3 Si 2 later, both dispersed in aluminum. The use of this technology requires a constant improvement of all processes involving the manufacture of MTR subject to several international protocols, which seek to ensure the reliability of the fuel from the standpoint of practical and environmental. In this context, the control of impurities, from the point of view of neutron economy, directly affects the quality of any nuclear fuel, so strict control is necessary. The literature has reported procedures which, beyond generating residues, are lengthy and costly, they need calibration curve and consequently reference materials. The aim of this work is to establish and validate a methodology for nondestructive quantitative chemical analysis, low cost and analysis time, as well as minimize the generation of waste, for multielement determination of major constituents (Utotal and Si) and impurities (B, Mg, Al, Cr, Mn, Fe, Co, Ni, Cu, Zn, Mo, Cd and others) present in U3O8 and U 3 Si 2 , meeting the needs of nuclear reactors in the nuclear fuel qualification type MTR. For that purposes, will be applied the X-ray fluorescence technique which allows fast chemical and nondestructive analysis, aside from sample preparation procedures that do not require previous chemical treatments (dissolving

  6. Power release estimation inside of fuel pins neighbouring fuel pin with gadolinium in a WWER-1000 type core

    International Nuclear Information System (INIS)

    Mikus, J.

    2006-01-01

    The purpose of this work consists in investigation of the gadolinium fuel pin (fps) influence on space power distribution, especially from viewpoint of the values and gradient occurrence inside of neighbouring FPs that could result in static loads with some consequences, e.g., FP bowing. Since detailed power distributions cannot be obtained in the NPPs, needed information is provided by means of experiments on research reactors. As for the power release measurement inside of FPs, some special (e.g. track) detectors placed between fuel pellets are usually used. Since such works are relatively complicated and time consuming, an evaluation method based on mathematical modelling and numerical approximation was proposed by means of that, and using measured (integral) power release in selected FPs, relevant information about power release inside of needed (investigated) FP, can be obtained. For this purpose, an experiment on light water, zero-power research reactor LR-0 was realized in a WWER-1000 type core with 7 fuel assemblies at zero boron concentration and containing gadolinium FPs. Application of the above evaluation method is demonstrated on investigated FP neighbouring a FP with gadolinium by means of the 1) Azimuthal power distribution inside of investigated FP on their fuel pellet surface in horizontal plane and 2) Gradient of the power distribution inside of investigated FP in two opposite positions on pellets surface that are situated to- and outwards a FP with gadolinium. Similar information can be relevant from the viewpoint of the FP failures occurrence investigation (Authors)

  7. Ammonia as a Suitable Fuel for Fuel Cells

    International Nuclear Information System (INIS)

    Lan, Rong; Tao, Shanwen

    2014-01-01

    Ammonia, an important basic chemical, is produced at a scale of 150 million tons per year. Half of hydrogen produced in chemical industry is used for ammonia production. Ammonia containing 17.5 wt% hydrogen is an ideal carbon-free fuel for fuel cells. Compared to hydrogen, ammonia has many advantages. In this mini-review, the suitability of ammonia as fuel for fuel cells, the development of different types of fuel cells using ammonia as the fuel and the potential applications of ammonia fuel cells are briefly reviewed.

  8. High performance liquid chromatographic hydrocarbon group-type analyses of mid-distillates employing fuel-derived fractions as standards

    Science.gov (United States)

    Seng, G. T.; Otterson, D. A.

    1983-01-01

    Two high performance liquid chromatographic (HPLC) methods have been developed for the determination of saturates, olefins and aromatics in petroleum and shale derived mid-distillate fuels. In one method the fuel to be analyzed is reacted with sulfuric acid, to remove a substantial portion of the aromatics, which provides a reacted fuel fraction for use in group type quantitation. The second involves the removal of a substantial portion of the saturates fraction from the HPLC system to permit the determination of olefin concentrations as low as 0.3 volume percent, and to improve the accuracy and precision of olefins determinations. Each method was evaluated using model compound mixtures and real fuel samples.

  9. Ammonia as a suitable fuel for fuel cells

    Directory of Open Access Journals (Sweden)

    Rong eLan

    2014-08-01

    Full Text Available Ammonia, an important basic chemical, is produced at a scale of 150 million tons per year. Half of hydrogen produced in chemical industry is used for ammonia production. Ammonia containing 17.5wt% hydrogen is an ideal carbon-free fuel for fuel cells. Compared to hydrogen, ammonia has many advantages. In this mini-review, the suitability of ammonia as fuel for fuel cells, the development of different types of fuel cells using ammonia as the fuel and the potential applications of ammonia fuel cells are briefly reviewed.

  10. Nuclear power plant types and the management of plutonium and minor actinides - in search of fuel cycle flexibility

    International Nuclear Information System (INIS)

    Thomas, J.B.

    2002-01-01

    Transuranics management concerns all NPP types, because of the specifications for sustainable development. Multiple recycling is mandatory. Neutronic abundance can be obtained in fast spectrum, or by adding external neutrons or (temporarily) with additional 235 U. The LWRs can control the plutonium inventory and significantly reduce the amount of transuranics transferred to the geological repository, thanks to the use of innovative nuclear fuel in a limited part of the NPP fleet. HTR adapted to transuranics burning can help. In the future, in addition to the liquid metal FBR, a strategy based on a gas cooled technological line and advanced fuel opens a second path towards fast spectra. Strategies for defining the optimal mix of reactor types in the nuclear fleet at a given time and demonstrating the fuel cycle flexibility are under study. (author)

  11. Study of homogeneous fuel cells type 10 x 10

    International Nuclear Information System (INIS)

    Montes, J.L.; Perusquia, R.; Ortiz, J.J.; Francois, J.L.; Marquez, C.M.

    2005-01-01

    At the moment in the National Institute of Nuclear Research (ININ) are carried out studies with the purpose of to establish a methodology that allows to carry out the neutron design of fuel cells of type 10 x 10. During the initial stage of the process of cells design, starting from the data that have to do with the planned energy demand it requires to be estimated the average value of the enrichment in U 235 w/o of the one assemble. The experience has shown that the accuracy that is achieved in this estimate it depends, among other factors, of the information (e.g. concentrations of U 235 and Gd 2 O 3 ) of the cells that its are disposed in that moment. For what we consider convenient to enlarge the available information by means of a series of calculations of cell physics; and to the one same time some aspects can be studied on the parameters that define the characteristics of a fuel cell. In this work the effect of the presence of different distributions of the concentrations of the fissile material is analyzed and of burnup poisons on the reactivity parameters of the cell as well as in the peak factor of local power (LPPF-Local Power Peaking Factor). (Author)

  12. HANARO fuel irradiation test(II)

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, D. S.; Kim, H. R.; Chae, H. T.; Lee, B. C.; Lee, C. S.; Kim, B. G.; Lee, C. B.; Hwang, W

    2001-04-01

    In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiatied at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%.

  13. Application of fire-retardant treatment to the wood in Type A unirradiated nuclear fuel outer containers

    International Nuclear Information System (INIS)

    Whitlow, J.D.; Luna, R.E.

    1992-01-01

    Packagings for transporting unirradiated nuclear fuel assemblies in the United States are commonly constructed as rectangular boxes consisting of a metal inner container, a wooden outer container, and cushioning material separating the two. The wood in the outer container is a potential source of fuel for fire. Use of a fire-retardant treatment on the wood may reduce or eliminate the damage to nuclear fuel assemblies in some types of accidents involving fire. The applicability of using fire-retardant treatments on the wood of outer containers is addressed. An approximate cost-benefit analysis to determine if fire-retardant treatments are economically justified is presented. (Author)

  14. Efetividade da sondagem pós-pilórica usando guia magnético Effectiveness of post-pyloric tube placement using magnetic guidance

    Directory of Open Access Journals (Sweden)

    Renata Andrea Pietro Pereira Viana

    2011-03-01

    Full Text Available OBJETIVOS: Suporte nutricional adequado tem papel importante na evolução de pacientes graves. Entretanto, significativa porcentagem destes pacientes evolui com dismotilidade intestinal, provocando alto volume gástrico residual. A administração de dieta enteral através de sonda em posição pós-pilórica tem sido sugerida como método para melhorar a tolerância. Objetivo deste estudo foi comparar a taxa de sucesso no posicionamento pós-pilórico da sonda nasoenteral por utilização de equipamento, que permite acompanhar a progressão da sonda através da visualização por transmissão eletromagnética em tempo real, em comparação com o método tradicional. MÉTODOS: Estudo prospectivo, randomizado, controlado, realizado em um hospital terciário durante três meses. Os pacientes foram randomizados para dois grupos: grupo com guia eletromagnético, pacientes submetidos à passagem de sonda nasoenteral sob auxilio do aparelho com visualização em tempo real e transmissão magnética e grupo convencional, passagem de sonda nasoenteral às cegas. O sucesso no posicionamento pós-pilórico e o tempo de duração do procedimento foram avaliados entre os grupos. RESULTADOS: Foram incluídos no estudo 37 pacientes, sendo 18 do grupo com guia eletromagnético e 19 do grupo convencional. A localização da sonda por meio da radiografia mostrou que o grupo com guia eletromagnético apresentou mais posicionamento pós-pilorico do que o grupo convencional, com menor tempo para realização do procedimento, com maior valor do pH do líquido aspirado pela sonda. CONCLUSÕES: O método de passagem e visualização a beira leito por transmissão eletromagnética garante de forma segura a monitorização e acurácia frente à sondagem nasoenteral.OBJECTIVE: Appropriate nutritional support is important to the outcomes of critically ill patients. However, a significant portion of these patients experience intestinal motility problems. Administration of

  15. Preliminary study of cost benefits associated with duplex fuel pellets of the LOWI type

    International Nuclear Information System (INIS)

    Ainscough, J.B.; Coucill, D.N.; Howl, D.A.; Jensen, A.; Misfeldt, I.

    1983-01-01

    Duplex UO 2 pellets, which consist of an outer enriched annulus and a depleted or natural core, can provide a solution to the problem of stress corrosion cracking failures, which have led to constraints being placed on ramp rates in power reactors. An analysis of the reactor physics and the performance of duplex pellets is presented in the context of a 17 X 17 pressurized water reactor fuel rod design. The study has been based on the particular type of duplex pellet in which the core and the annulus are physically separate; this is called ''LOWI'' after the Danish design. At low burnup, this fuel shows a significant improvement in power ramp performance compared with standard fuel. At higher burnup, the benefits are less certain but as the severity of the ramp will usually be less in high burnup fuel simply because of the reduced rating, the reduction in benefit may not be significant. If the gap between the core and annulus persists to high burnup, there will be no loss of benefit. Economic calculations and a cost-benefit analysis are presented to show the number of extra full-power hours of reactor operation that must be obtained in order to outweigh the additional fabrication costs associated with this fuel

  16. Analytical Evaluation to Determine Selected PAHs by HPLC in a Type 2 Fuel; Evaluacion Analitica de 4 Metodos de Determinacion de PAHs medianteHPLC en un Fuel de Tipo II

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Escolano Segovia, O.; Garcia Frutos, F. J.

    2009-05-21

    An evaluation of analytical parameters to determine selected PAHs in a fuel oil type II by HPLC coupled to fluorescence and diode detectors is presented. The study was focused on four conventional treatments of these kinds of oil samples and the main objective was giving a measure of confidence level of PAH results in the fuel oil. This study was performed in the frame of the project Assessment of natural attenuation of PAHs in agricultural soil contaminated with fuel from an accidental spill (Spanish National Plain I+D+I, CTM2007-64537). This paper is presented as follows: Analysis of reference material 1582 (NIST) by using the four kinds of sample treatments of interest. Application of variance analysis to compare results obtained from type II fuel by using each sample treatment and chromatographic detector. Finally, a statistic calculation was performed to measure uncertainty components in chromatographic analysis. (Author)

  17. Experimental study on the 300W class planar type solid oxide fuel cell stack: Investigation for appropriate fuel provision control and the transient capability of the cell performance

    International Nuclear Information System (INIS)

    Komatsu, Y; Brus, G; Szmyd, J S; Kimijima, S

    2012-01-01

    The present paper reports the experimental study on the dynamic behavior of a solid oxide fuel cell (SOFC). The cell stack consists of planar type cells with standard power output 300W. A Major subject of the present study is characterization of the transient response to the electric current change, assuming load-following operation. The present studies particularly focus on fuel provision control to the load change. Optimized fuel provision improves power generation efficiency. However, the capability of SOFC must be restricted by a few operative parameters. Fuel utilization factor, which is defined as the ratio of the consumed fuel to the supplied fuel is adopted for a reference in the control scheme. The fuel flow rate was regulated to keep the fuel utilization at 50%, 60% and 70% during the current ramping. Lower voltage was observed with the higher fuel utilization, but achieved efficiency was higher. The appropriate mass flow control is required not to violate the voltage transient behavior. Appropriate fuel flow manipulation can contribute to moderate the overshoot on the voltage that may appear to the current change. The overshoot on the voltage response resulted from the gradual temperature behavior in the SOFC stack module.

  18. Experimental study on the 300W class planar type solid oxide fuel cell stack: Investigation for appropriate fuel provision control and the transient capability of the cell performance

    Science.gov (United States)

    Komatsu, Y.; Brus, G.; Kimijima, S.; Szmyd, J. S.

    2012-11-01

    The present paper reports the experimental study on the dynamic behavior of a solid oxide fuel cell (SOFC). The cell stack consists of planar type cells with standard power output 300W. A Major subject of the present study is characterization of the transient response to the electric current change, assuming load-following operation. The present studies particularly focus on fuel provision control to the load change. Optimized fuel provision improves power generation efficiency. However, the capability of SOFC must be restricted by a few operative parameters. Fuel utilization factor, which is defined as the ratio of the consumed fuel to the supplied fuel is adopted for a reference in the control scheme. The fuel flow rate was regulated to keep the fuel utilization at 50%, 60% and 70% during the current ramping. Lower voltage was observed with the higher fuel utilization, but achieved efficiency was higher. The appropriate mass flow control is required not to violate the voltage transient behavior. Appropriate fuel flow manipulation can contribute to moderate the overshoot on the voltage that may appear to the current change. The overshoot on the voltage response resulted from the gradual temperature behavior in the SOFC stack module.

  19. Development of the Simulation Program for the In-Vessel Fuel Handling System of Double Rotating Plug Type

    International Nuclear Information System (INIS)

    Kim, S. H.; Kim, J. B.

    2011-01-01

    In-vessel fuel handling machines are the main equipment of the in-vessel fuel handling system, which can move the core assembly inside the reactor vessel along with the rotating plug during refueling. The in vessel fuel handling machines for an advanced sodium cooled fast reactor(SFR) demonstration plant are composed of a direct lift machine(DM) and a fixed arm machine(FM). These machines should be able to access all areas above the reactor core by means of the rotating combination of double rotating plugs. Thus, in the in vessel fuel handling system of the double rotating plug type, it is necessary to decide the rotating plug size and evaluate the accessibility of in-vessel fuel handling machines in given core configuration. In this study, the simulation program based on LABVIEW which can effectively perform the arrangement design of the in vessel fuel handling system and simulate the rotating plug motion was developed. Fig. 1 shows the flow chart of the simulation program

  20. Fuel pellet relocation behavior in fast reactor uranium-plutonium mixed oxide fuel pin at beginning-of-life

    International Nuclear Information System (INIS)

    Inoue, Masaki; Ukai, Shigeharu; Asaga, Takeo

    1999-08-01

    The effects of fabrication parameters, irradiation conditions and fuel microstructural feature on fuel pellet relocation behavior in fast reactor fuel pins were investigated. This work focused only on beginning-of-life conditions, when fuel centerline temperature depends largely on the behavior. Fuel pellet relocation behavior in Joyo Mk-II driver could not be characterized because of the lack of data. And the behavior in FFTF driver and its larger diameter type fuel pins could not be characterized because of the extensive lot-by-lot scatters. The behavior both in Monju type and in Joyo power-to-melt type fuel pins were similar to each other, and depends largely on the as-fabricated gap width while the effects of linear heat rate and the extent of microstructural evolution were negligible. And fuel pellet centerline melting seems to affect slightly the behavior. The correlation, which describes the extent of relocation both in Monju type and in Joyo power-to-melt type fuel pins, were newly formulated and extrapolated for Joyo Mk-II driver, FFTF driver and its larger diameter type fuel pins. And the behavior in Joyo Mk-II driver seemed to be similar. On the contrary, the similarity with JNC fuel pins was observed case-by-case in FFTF driver and its larger diameter type fuel pins. (author)

  1. Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

    OpenAIRE

    Itamar Iliuk; José Manoel Balthazar; Ângelo Marcelo Tusset; José Roberto Castilho Piqueira

    2016-01-01

    Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was prop...

  2. Fuels processing for transportation fuel cell systems

    Science.gov (United States)

    Kumar, R.; Ahmed, S.

    Fuel cells primarily use hydrogen as the fuel. This hydrogen must be produced from other fuels such as natural gas or methanol. The fuel processor requirements are affected by the fuel to be converted, the type of fuel cell to be supplied, and the fuel cell application. The conventional fuel processing technology has been reexamined to determine how it must be adapted for use in demanding applications such as transportation. The two major fuel conversion processes are steam reforming and partial oxidation reforming. The former is established practice for stationary applications; the latter offers certain advantages for mobile systems and is presently in various stages of development. This paper discusses these fuel processing technologies and the more recent developments for fuel cell systems used in transportation. The need for new materials in fuels processing, particularly in the area of reforming catalysis and hydrogen purification, is discussed.

  3. Nuclear fuel burnup calculation in a Voronezh type reactor; Analiza izgaranja nuklearnog goriva u reaktoru tipa Voronjez

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M; Marinkovic, N; Kocic, A [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1977-07-01

    In order to summarize and present our abilities to perform a complex computation of the nuclear fuel burn-up, a systematic review of the available methods, algorithms and computer programmes is given in this paper. The computer programmes quoted have all been developed, modified and tested in our department, so that they can be successfully used in the analysis of nuclear power plants from both physics and economic points of view. For a commercially proven nuclear reactor - reactor of the Voronezh type - an illustrative computation of the fuel burn-up is performed. The typical results are presented and discussed. The conclusion concerns the completion of a modular scheme for the fuel burn-up calculation and the fuel cycle analysis (author)

  4. The technique for determination of surface contamination by uranium on U3Si2-Al plate-type fuel elements

    International Nuclear Information System (INIS)

    Li Shulan; He Fengqi; Wang Qingheng; Han Jingquan

    1993-04-01

    The NDT method for determining the surface contamination by uranium on U 3 Si 2 -Al plate-type fuel elements, the process of standard specimen preparation and the graduation curve are described. The measurement results of U 3 Si 2 -Al plate-type fuel elements show that the alpha counting method to measure the surface contamination by uranium on fuel plate is more reliable. The UB-1 type surface contamination meter, which was recently developed, has many advantages such as high sensitivity to determine the uranium pollution, short time in measuring, convenience for operation, and the minimum detectable amount of uranium is 5 x 10 -10 g/cm 2 . The measuring device is controlled by a microcomputer. Besides data acquisition and processing, it has functions of statistics, output data on terminal or to printer and alarm. The procedures of measurement are fully automatic. All of these will meet the measuring needs in batch process

  5. A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Makmal, T. [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel); Nuclear Physics and Engineering Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Aviv, O. [Radiation Safety Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Gilad, E., E-mail: gilade@bgu.ac.il [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel)

    2016-10-21

    A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections. - Highlights: • Simple, inexpensive, safe and flexible experimental setup that can be quickly deployed. • Experimental results are thoroughly corroborated against ORIGEN2 burnup code. • Experimental uncertainty of 9% and 5% deviation between measurements and simulations. • Very high burnup MTR fuel element is examined, with 60% depletion of {sup 235}U. • Impact of highly irregular irradiation regime on burnup evaluation is studied.

  6. Safety characteristics of mid-sized MOX fueled liquid metal reactor core of high converter type in the initiating phase of unprotected loss of flow accident. Effect of low specific fuel power density on ULOF behavior brought by employment of large diameter fuel pins

    International Nuclear Information System (INIS)

    Ishida, Masayoshi; Kawada, Kenichi; Niwa, Hajime

    2003-07-01

    Safety characteristics in core disruptive accidents (CDAs) of mid-sized MOX fueled liquid metal reactor core of high converter type have been examined by using the CDA initiating phase analysis code SAS4A. The design concept of high converter type reactor core has been studied as one of options in the category of sodium-cooled reactor in Phase II of Feasibility Study on Commercialized Fast Reactor Cycle System. An unprotected loss-of-flow accident (ULOF) has been selected as a representative CDA initiator for this study. A core concept of high converter type, which employed a large diameter fuel pin of 11.1 mm with 1.2 m core height to get a large fuel volume fraction in the core to achieve high internal conversion ratio was proposed in JFY2001. Each fuel subassembly of the core (abbreviated here as UPL120)was provided with an upper sodium plenum directly above the core to reduce the sodium void reactivity worth. Because of the large fuel pin diameter, average specific fuel power density (31 kW/kg-MOX) of UPL120 is about one half of those of conventional large MOX cores. The reactivity worth of sodium voiding is 6$ in the whole core, and -1$ in the all upper plenums. Initiating phase of ULOF accident in UPL120 under the conditions of nominal design and best estimate analysis resulted in a slightly super-prompt critical power burst. The causes of the super-prompt criticality have been identified twofold: (a) the low specific fuel power density of core reduced the effectiveness of prompt negative reactivity feedback of Doppler and axial fuel expansion effects upon increase in reactor power, and (b) the longer core height compared with conventional 1m cores brought, together with the lower specific power density, a remarkable delay in insertion of negative fuel dispersion reactivity after the onset of fuel disruption in sodium voided subassembly due to the lower linear heat rating in the top portion of the core. During the delay, burst-type fuel failures in sodium un

  7. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 1. Summary: alternatives for the back of the LWR fuel cycle types and properties of LWR fuel cycle wastes projections of waste quantities; selected glossary

    International Nuclear Information System (INIS)

    1976-05-01

    Volume I of the five-volume report contains executive and technical summaries of the entire report, background information of the LWR fuel cycle alternatives, descriptions of waste types, and projections of waste quantities. Overview characterizations of alternative LWR fuel cycle modes are also included

  8. Metal membrane-type 25-kW methanol fuel processor for fuel-cell hybrid vehicle

    Science.gov (United States)

    Han, Jaesung; Lee, Seok-Min; Chang, Hyuksang

    A 25-kW on-board methanol fuel processor has been developed. It consists of a methanol steam reformer, which converts methanol to hydrogen-rich gas mixture, and two metal membrane modules, which clean-up the gas mixture to high-purity hydrogen. It produces hydrogen at rates up to 25 N m 3/h and the purity of the product hydrogen is over 99.9995% with a CO content of less than 1 ppm. In this fuel processor, the operating condition of the reformer and the metal membrane modules is nearly the same, so that operation is simple and the overall system construction is compact by eliminating the extensive temperature control of the intermediate gas streams. The recovery of hydrogen in the metal membrane units is maintained at 70-75% by the control of the pressure in the system, and the remaining 25-30% hydrogen is recycled to a catalytic combustion zone to supply heat for the methanol steam-reforming reaction. The thermal efficiency of the fuel processor is about 75% and the inlet air pressure is as low as 4 psi. The fuel processor is currently being integrated with 25-kW polymer electrolyte membrane fuel-cell (PEMFC) stack developed by the Hyundai Motor Company. The stack exhibits the same performance as those with pure hydrogen, which proves that the maximum power output as well as the minimum stack degradation is possible with this fuel processor. This fuel-cell 'engine' is to be installed in a hybrid passenger vehicle for road testing.

  9. Investigation into the effect of different fuels on ignition delay of M-type diesel combustion process

    Directory of Open Access Journals (Sweden)

    Bibić Dževad

    2008-01-01

    Full Text Available An ignition delay is a very complex process which depends on a great number of parameters. In practice, definition of the ignition delay is based on the use of correlation expressions. However, the correlation expressions have very often limited application field. This paper presents a new correlation which has been developed during the research project on the direct injection M-type diesel engine using both the diesel and biodiesel fuel, as well as different values of a static injection timing. A dynamic start of injection, as well as the ignition delay, is defined in two ways. The first approach is based on measurement of a needle lift, while the second is based on measurement of a fuel pressure before the injector. The latter approach requires calculation of pressure signals delay through the fuel injection system and the variation of a static advance injection angle changing. The start of a combustion and the end of the ignition delay is defined on the basis of measurements of an in-cylinder pressure and its point of separation from a skip-fire pressure trace. The developed correlation gives better prediction of the ignition delay definition for the M-type direct injection diesel engine in the case of diesel and biodiesel fuel use when compared with the classic expression by the other authors available in the literature.

  10. Fuel cell systems

    International Nuclear Information System (INIS)

    Kotevski, Darko

    2003-01-01

    Fuel cell systems are an entirely different approach to the production of electricity than traditional technologies. They are similar to the batteries in that both produce direct current through electrochemical process. There are six types of fuel cells each with a different type of electrolyte, but they all share certain important characteristics: high electrical efficiency, low environmental impact and fuel flexibility. Fuel cells serve a variety of applications: stationary power plants, transport vehicles and portable power. That is why world wide efforts are addressed to improvement of this technology. (Original)

  11. Utilization of the experimental reactor Osiris for the study and the development of fuels of the fast neutron reactor type

    International Nuclear Information System (INIS)

    Marcon, M.; Faugere, J.L.; Genthon, J.P.; Maillot, R.

    1977-01-01

    Nuclear fuel tests for the fast neutron reactor type have been carried out at the Osiris reactor: thermal study of (U,Pu)O 2 oxide by measurement with thermocouples in the core of the fuel pellet; study of the effects of power cycling on nuclear fuel; study of the mechanical interactions between oxide and cladding by measurement of the cladding deformation during irradiation [fr

  12. Dilatational behaviour of ZrNb1 fuel cans of a WWER-type reactor during a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Adam, E.; Stephan, M.; Wetzel, L.

    1987-01-01

    Based on an assessment of various factors of influence on the performance of fuel cans during normal operation and imaginable accidents, the necessity of studying creep and burst behaviour of WWER-type fuel cans of ZrNb1 under simulated LOCA conditions has been proved and an experimental facility designed for this purpose is described. Control of fuel can temperature is accomplished through a minicomputer during the creep and bursts experiments. With this, various temperature loading profiles of the fuel cans can be realized. Experimental results on dilatational behaviour of ZrNb1 fuel cans from isothermal creep and burst experiments in air are presented and compared with values for Zircaloy. (author)

  13. Fuel pin behavior under slow ramp-type transient-overpower conditions in the CABRI-FAST experiments

    International Nuclear Information System (INIS)

    Fukano, Yoshitaka; Onoda, Yuichi; Sato, Ikken; Charpenel, Jean

    2009-01-01

    In the CABRI-FAST experimental program, four in-pile tests were performed with slow power-ramp-type transient-overpower conditions (called hereafter as 'slow TOP') to study transient fuel pin behavior under inadvertent control rod withdrawal events in liquid metal cooled fast breeder reactors. Annular-pellet fuel pins were used in three tests, while a solid-pellet fuel pin was used in the other test. All of these pins were pre-irradiated in Phenix. The slow TOP test with a solid-pellet fuel pin was realized as a comparatory test against an existing test (E12) in the CABRI-2 program. In the CABRI-FAST test (BCF1), a power ramp rate of 3% Po/s was applied, while in the CABRI-2 test, 1% Po/s was adopted. Moreover, overpower condition was maintained for a few seconds beyond the observed pin failure in the BCF1 test. In spite of the different power ramp rates, evaluated fuel thermal conditions at the observed failure time are quite similar. The continued overpower condition in the BCF1 test resulted in gradual degradation of the pin structure providing information effective for evaluation of various accident scenarios. Three slow TOP tests with the annular fuel in the CABRI-FAST program resulted in no pin failure showing high failure threshold. Based on post-test examination data and a theoretical evaluation, it was concluded that intra-pin free spaces, such as central hole, macroscopic cracks and fuel-cladding gap effectively mitigated fuel cladding mechanical interaction. It was also clarified that cavity pressurization became effective only in case of very large amount of fuel melting. Furthermore, such cavity pressurization was effectively mitigated by a molten-fuel squirting into the upper blanket region pushing the blanket pellets upward. These CABRI FAST slow TOP tests, in combination with the existing CABRI and TREAT tests, provided an extended slow TOP test database with various fuel and transient conditions. (author)

  14. HTPEM Fuel Cell Impedance

    DEFF Research Database (Denmark)

    Vang, Jakob Rabjerg

    As part of the process to create a fossil free Denmark by 2050, there is a need for the development of new energy technologies with higher efficiencies than the current technologies. Fuel cells, that can generate electricity at higher efficiencies than conventional combustion engines, can...... potentially play an important role in the energy system of the future. One of the fuel cell technologies, that receives much attention from the Danish scientific community is high temperature proton exchange membrane (HTPEM) fuel cells based on polybenzimidazole (PBI) with phosphoric acid as proton conductor....... This type of fuel cell operates at higher temperature than comparable fuel cell types and they distinguish themselves by high CO tolerance. Platinum based catalysts have their efficiency reduced by CO and the effect is more pronounced at low temperature. This Ph.D. Thesis investigates this type of fuel...

  15. Calculation of the linear heat generation rates which violate the thermomechanical limit of plastic deformation of the fuel cladding in function of the burn up of a BWR fuel rod type

    International Nuclear Information System (INIS)

    Lucatero, M.A.; Hernandez L, H.

    2003-01-01

    The linear heat generation rates (LHGR) for a BWR type generic fuel rod, as function of the burnup that violate the thermomechanical limit of circumferential plastic deformation of the can (canning) in nominal operation in stationary state of the fuel rod are calculated. The evaluation of the LHGR in function of the burnt of the fuel, is carried out under the condition that the deformation values of the circumferential plastic deformation of the can exceeds in 0.1 the thermomechanical value operation limit of 1%. The results of the calculations are compared with the generation rates of linear operation heat in function of the burnt for this fuel rod type. The calculations are carried out with the FEMAXI-V and RODBURN codes. The results show that for exhibitions or burnt between 0 and 16,000 M Wd/tU a minimum margin of 160.8 W/cm exists among LHGR (439.6 W/cm) operation peak for the given fuel and maximum LHGR of the fuel (calculated) to reach 1.1% of circumferential plastic deformation of the can, for the peak factor of power of 1.40. For burnt of 20,000 MWd/tU and 60,000 MWd/tU exist a margin of 150.3 and 298.6 W/cm, respectively. (Author)

  16. Control console conceptual design for sheet type fuels of Triga Mark-II reactor

    International Nuclear Information System (INIS)

    Eko Priyono; Kurnia Wibowo; Anang Susanto

    2016-01-01

    The control console conceptual design for sheet type fuel of TRIGA Mark-II reactor has been made. The control console conceptual design was made with refer study result of instrument and control system which is used in BATAN'S reactor i.e TRIGA-2000 Bandung, TRIGA Yogyakarta and MPR-30 Serpong. The control console conceptual design was made by using AutoCad software. The control console conceptual design reactor for sheet type fuel of TRIGA Mark-II reactor consist of 5 segments that is 3 segments for placing the computer monitors, 1 segment for placing bargraph displays and recorders and 1 segment for placing panel meters. There are the door on front and back position at each segment for enter and out devices in the console. The control console conceptual design is also equipped by the table along in front of console for placing reactor panel control and for writing, 3 drawers for 3 keyboards. The dimension of console will refer control room size and the components will be placed on console which will be detailed in detail design if this conceptual design has been approved. (author)

  17. Performance enhancement of a spark ignition engine fed by different fuel types

    International Nuclear Information System (INIS)

    Hedfi, Hachem; Jbara, Abdessalem; Jedli, Hedi; Slimi, Khalifa; Stoppato, Anna

    2016-01-01

    Highlights: • Biogas mixed with hydrogen is checked for a spark ignition engine. • An engine fed by biogas, hydrogen, natural gas or liquid petroleum gas is studied. • Efficiency is optimized with respect to consumption and exhaust gas recirculation. • Combustion reaction progress is characterized in real time. - Abstract: A numerical model based on thermodynamic and kinetic analyses has been established in order to evaluate biogas, hydrogen, natural gas or liquid petroleum gas as fuels in a spark ignition engine. For each fuel type, consumption as well as efficiency have been compared to gasoline in order to generate the same engine work (in the range of 0.28–0.43 W h/cycle). It was found that the spark ignition engine can be fed by an equimolar mixture of biogas and hydrogen. Moreover, thermal efficiency has been enhanced with respect to fuel consumption and exhaust gas recirculation (EGR). It was shown that an equimolar mixture between biogas and hydrogen increases the ITE by around 2.2% and decreases the mass consumption by less than 0.01 g/cycle. In addition, the combustion reaction progresses as well as CO and CO_2 emissions have been characterized in real time.

  18. Concept of safe tank-type water cooled and moderated reactor with HTGR microparticle fuel compacts

    International Nuclear Information System (INIS)

    Gol'tsev, A.O.; Kukharkin, N.E.; Mosevitskij, I.S.; Ponomarev-Stepnoj, N.N.; Popov, S.V.; Udyanskij, Yu.N.; Tsibul'skij, V.F.

    1993-01-01

    Concept of safe tank-type water-cooled and moderated reactor on the basis of HTGR fuel microparticles which enable to avoid environment contamination with radioactive products under severe accidents, is proposed. Results of neutron-physical and thermal-physical studies of water cooled and moderated reactor with HTGR microparticle compacts are presented. Characteristics of two reactors with thermal power of 500 and 1500 MW are indicated within the concept frames. The reactor behaviour under severe accident connected with complete loss of water coolant is considered. It is shown that under such an accident the fission products release from fuel microparticles does not occur

  19. Evidence of asymmetric behavioral responses to changes in gasoline prices and taxes for different fuel types

    International Nuclear Information System (INIS)

    Bajo-Buenestado, Raúl

    2016-01-01

    Using monthly data from the Spanish gasoline retail market we explore asymmetries in consumers’ behavioral responses to changes in gasoline prices and taxes. In particular, we are interested in investigating whether an increase in gasoline taxes has a more negative impact on the demand than a –similar in magnitude– increase in the “pre-tax” price of gasoline for different fuel types. We estimate fuel consumers’ responses using a rich set of robust panel data models considering potential dynamic effects and endogeneity problems. We find evidence to confirm the existence of asymmetric responses for the demand of unleaded fuels and agricultural diesel fuel. However we cannot support this statement for the regular diesel case: for this fuel both the tax-exclusive price and the tax elasticities are roughly the same. This result agrees with the fact that “diesel drivers” tend to be better informed about changes in both fuel prices and taxes. Some implications in terms of fiscal policy and pollution and climate change policy are also discussed. - Highlights: •Provide evidence of asymmetric responses of gasoline demand due to changes in prices and taxes. •Identify differences in the elasticity of the demand of diesel fuel and unleaded gasoline. •Perform robustness checks considering dynamic effects and IV regression. •Provide some policy recommendations for future gasoline tax changes.

  20. Fuel fabrication and reprocessing for nuclear fuel cycle with inherent safety demands

    Energy Technology Data Exchange (ETDEWEB)

    Shadrin, Andrey Yurevich; Dvoeglazov, Konstantin Nikolaevich; Ivanov, Valentine Borisovich; Volk, Vladimir Ivanovich; Skupov, Mikhail Vladimirovich; Glushenkov, Alexey Evgenevich [Joint Stock Company ' ' The High Technological Research Institute of Inorganic Materials' ' , Moscow (Russian Federation); Troyanov, Vladimir Mihaylovich; Zherebtsov, Alexander Anatolievich [Innovation and Technology Center of Project ' ' PRORYV' ' , State Atomic Energy Corporation ' ' Rosatom' ' , Moscow (Russian Federation)

    2015-06-01

    The strategies adopted in Russia for a closed nuclear fuel cycle with fast reactors (FR), selection of fuel type and recycling technologies of spent nuclear fuel (SNF) are discussed. It is shown that one of the possible technological solutions for the closing of a fuel cycle could be the combination of pyroelectrochemical and hydrometallurgical methods of recycling of SNF. This combined scheme allows: recycling of SNF from FR with high burn-up and short cooling time; decreasing the volume of stored SNF and the amount of plutonium in a closed fuel cycle in FR; recycling of any type of SNF from FR; obtaining the high pure end uranium-plutonium-neptunium end-product for fuel refabrication using pellet technology.

  1. Comparison of the parameters of the IR-8 reactor with different fuel assembly designs with LEU fuel

    International Nuclear Information System (INIS)

    Vatulin, A.; Stetsky, Y.; Dobrikova, I.

    1999-01-01

    The estimation of neutron-physical, heat and hydraulic parameters of the IR-8 research reactor with low enriched uranium (LEU) fuel was performed. Two fuel assembly (FA) designs were reviewed: IRT-4M with the tubular type fuel elements and IRT-MR with the rod type fuel elements. UO 2 -Al dispersion 19.75% enrichment fuel is used in both cases. The results of the calculations were compared with main parameters of the reactor, using the current IRT-3M FA with 90% high enriched uranium (HEU) fuel. The results of these comparisons showed that during the LEU conversion of the reactor the cycle length, excess reactivity and peak power of the IRT-MR type FA are higher than for the IRT-3M type FA and IRT-4M type FA. (author)

  2. HANARO fuel irradiation test (II): revision

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, D. S.; Kim, H.; Chae, H. T.; Lee, C. S.; Kim, B. G.; Lee, C. B

    2001-04-01

    In order to fulfill the requirement to prove HANARO fuel integrity when irradiated at a power greater than 112.8 kW/m, which was imposed during HANARO licensing, and to verify the irradiation performance of HANARO fuel, the in-pile irradiation test of HANARO fuel has been performed. Two types of test fuel, the un-instrumented Type A fuel for higher burnup irradiation in shorter period than the driver fuel and the instrumented Type B fuel for higher linear heat rate and precise measurement of irradiation conditions, have been designed and fabricated. The test fuel assemblies were irradiated in HANARO. The two Type A fuel assemblies were intended to be irradiated to medium and high burnup and have been discharged after 69.9 at% and 85.5 at% peak burnup, respectively. Type B fuel assembly was intended to be irradiated at high power with different instrumentations and achieved a maximum power higher than 120 kW/m without losing its integrity and without showing any irregular behavior. The Type A fuel assemblies were cooled for about 6 months and transported to the IMEF(Irradiated Material Examination Facility) for consequent evaluation. Detailed non-destructive and destructive PIE (Post-Irradiation Examination), such as the measurement of burnup distribution, fuel swelling, clad corrosion, dimensional changes, fuel rod bending strength, micro-structure, etc., has been performed. The measured results have been analysed/compared with the predicted performance values and the design criteria. It has been verified that HANARO fuel maintains proper in-pile performance and integrity even at the high power of 120 kw/m up to the high burnup of 85 at%. This report is the revision of KAERI/TR-1816/2001 on the irradiation test for HANARO fuel.

  3. Computational comparison of the effect of mixing grids of 'swirler' and 'run-through' types on flow parameters and the behavior of steam phase in WWER fuel assemblies

    International Nuclear Information System (INIS)

    Shcherbakov, S.; Sergeev, V.

    2011-01-01

    The results obtained using the TURBOFLOW computer code are presented for the numerical calculations of space distributions of coolant flow, heating and boiling characteristics in WWER fuel assemblies with regard to the effect of mixing grids of 'Swirler' and 'Run-through' types installed in FA on the above processes. The nature of the effect of these grids on coolant flow was demonstrated to be different. Thus, the relaxation length of cross flows after passing a 'Run-through' grid is five times as compared to a 'Swirler'-type grid, which correlates well with the experimental data. At the same time, accelerations occurring in the flow downstream of a 'Swirler'-type grid are by an order of magnitude greater than those after a 'Run-through' grid. As a result, the efficiency of one-phase coolant mixing is much higher for the grids of 'Run-through' type, while the efficiency of steam removal from fuel surface is much higher for 'Swirler'-type grids. To achieve optimal removal of steam from fuel surface it has been proposed to install into fuel assemblies two 'Swirler'-type grids in tandem at a distance of about 10 cm from each other with flow swirling in opposite directions. 'Run-through' grids would be appropriate for use for mixing in fuel assemblies with a high non-uniformity of fuel-by-fuel power generation. (authors)

  4. Characterization of dispersed type fuel miniplates based in alloy UMo by evaluation of changes volumetrics and thermal conductivity

    International Nuclear Information System (INIS)

    Salinas Valero, Pablo Ignacio

    2016-01-01

    The development of new technologies in the nuclear area is extremely important to achieve greater efficiency and security in the production of electrical energy in the case of power reactors and for the production of radioisotopes and neutrons in research reactors. Throughout history, uranium-based nuclear fuels evolved in parallel with the requirements of nuclear reactors, this emphasis was increased when the RERTR program was created, which restricts the use of fuels with a maximum enrichment of 20% of the isotope U 235 (fissile isotope), which makes it necessary to increase the mass of uranium to compensate the amount of fissile material to maintain a neutron flux necessary for the reactors to operate with the same power. The search for new nuclear fuels has reached the UMo alloy with which densities of 18 gU/cm 3 are achieved in type fuels and 8 gU/cm 3 in dispersed type fuels, properties under irradiation due to their cubic crystalline structure. This type of fuel, when used dispersed in an aluminum matrix, becomes thermodynamically unstable by increasing the fission temperature of the U 235 isotope, due to this, compounds of lower density are formed, which causes an increase in volume (swelling). ). This swelling is studied throughout the present work, to relate the changes of UMo-Al / 4% volume of thermally induced miniecography in thermal treatments, with the purpose of evaluating changes in the thermal conductivity of the material. In this study it was detected that the swelling in miniplates is related in some way to the reduction of thermal conductivity, it was also recorded that the volume of change is irregular increasing and decreasing its volume according to the hours of induced swelling. The purpose of this work is to contribute to the development of dispersed fuels based on the UMo alloy in order to control the variables and reduce the probability of faults and possible accidents, such as fission products, or an increase in temperature in the core

  5. Os rankings do Guia do Estudante na educação superior brasileira: um estudo sobre as estratégias de divulgação adotadas pelas instituições que obtiveram o Prêmio Melhores Universidades

    OpenAIRE

    Lourenço, Henrique da Silva

    2014-01-01

    A presente dissertação se enquadra no campo das pesquisas em avaliação educacional e tem foco no estudo das classificações advindas do Guia do Estudante (GE), publicada pela Editora Abril. A escassez de pesquisas voltadas ao estudo da problemática torna relevante a presente investigação, na medida em que se pretende compreender como ocorre o processo de comunicação dos resultados obtidos pelas universidades brasileiras públicas e particulares (com ou sem fins lucrativos) no Prêmio Melhore...

  6. Hydrodeoxygenation of oxidized distilled bio-oil for the production of gasoline fuel type

    International Nuclear Information System (INIS)

    Luo, Yan; Guda, Vamshi Krishna; Hassan, El Barbary; Steele, Philip H.; Mitchell, Brian; Yu, Fei

    2016-01-01

    Highlights: • Oxidation had more influence on the yield of total hydrocarbons than distillation. • The highest total hydrocarbon yield was obtained from oxidized distilled bio-oil. • The 2nd-stage hydrocarbons were in the range of gasoline fuel boiling points. • The main products for upgrading of oxidized bio-oil were aliphatic hydrocarbons. • The main products for upgrading of non-oxidized bio-oil were aromatic hydrocarbons. - Abstract: Distilled and oxidized distilled bio-oils were subjected to 1st-stage mild hydrodeoxygenation and 2nd-stage full hydrodeoxygenation using nickel/silica–alumina catalyst as a means to enhance hydrocarbon yield. Raw bio-oil was treated for hydrodeoxygenation as a control to which to compare study treatments. Following two-stage hydrodeoxygenation, four types of hydrocarbons were mainly comprised of gasoline and had water contents, oxygen contents and total acid numbers of nearly zero and higher heating values of 44–45 MJ/kg. Total hydrocarbon yields for raw bio-oil, oxidized raw bio-oil, distilled bio-oil and oxidized distilled bio-oil were 11.6, 16.2, 12.9 and 20.5 wt.%, respectively. The results indicated that oxidation had the most influence on increasing the yield of gasoline fuel type followed by distillation. Gas chromatography/mass spectrometry characterization showed that 66.0–76.6% of aliphatic hydrocarbons and 19.5–31.6% of aromatic hydrocarbons were the main products for oxidized bio-oils while 35.5–38.7% of aliphatic hydrocarbons and 58.2–63.1% of aromatic hydrocarbons were the main products for non-oxidized bio-oils. Both aliphatic and aromatic hydrocarbons are important components for liquid transportation fuels and chemical products.

  7. Fuel assemblies for use in FBR type reactors

    International Nuclear Information System (INIS)

    Nakahigashi, Shigeo; Terasawa, Michitaka.

    1984-01-01

    Purpose: To prevent slackings in lapping wires and thereby enabling to maintain the distance between fuel pins always constant during use. Constitution: Lapping wires are wound helically around the outer circumference of each fuel pin in order to maintain the distance between fuel pins constant and unify the flow of coolants. The material of the lapping wire is defined as below. Specifically, austenite stainless steels incorporated with 0.045% titanium are used in the state of molten procession material as they are without no further cold working. Lapping wires having anti-swelling property can be obtained with this material and the slackings in the lapping wires during use can be prevented. (Ikeda, J.)

  8. Commuters’ Exposure to Particulate Matter Air Pollution Is Affected by Mode of Transport, Fuel Type, and Route

    Science.gov (United States)

    Zuurbier, Moniek; Hoek, Gerard; Oldenwening, Marieke; Lenters, Virissa; Meliefste, Kees; van den Hazel, Peter; Brunekreef, Bert

    2010-01-01

    Background Commuters are exposed to high concentrations of air pollutants, but little quantitative information is currently available on differences in exposure between different modes of transport, routes, and fuel types. Objectives The aim of our study was to assess differences in commuters’ exposure to traffic-related air pollution related to transport mode, route, and fuel type. Methods We measured particle number counts (PNCs) and concentrations of PM2.5 (particulate matter ≤ 2.5 μm in aerodynamic diameter), PM10, and soot between June 2007 and June 2008 on 47 weekdays, from 0800 to 1000 hours, in diesel and electric buses, gasoline- and diesel-fueled cars, and along two bicycle routes with different traffic intensities in Arnhem, the Netherlands. In addition, each-day measurements were taken at an urban background location. Results We found that median PNC exposures were highest in diesel buses (38,500 particles/cm3) and for cyclists along the high-traffic intensity route (46,600 particles/cm3) and lowest in electric buses (29,200 particles/cm3). Median PM10 exposure was highest from diesel buses (47 μg/m3) and lowest along the high- and low-traffic bicycle routes (39 and 37 μg/m3). The median soot exposure was highest in gasoline-fueled cars (9.0 × 10−5/m), diesel cars (7.9 × 10−5/m), and diesel buses (7.4 × 10−5/m) and lowest along the low-traffic bicycle route (4.9 × 10−5/m). Because the minute ventilation (volume of air per minute) of cyclists, which we estimated from measured heart rates, was twice the minute ventilation of car and bus passengers, we calculated that the inhaled air pollution doses were highest for cyclists. With the exception of PM10, we found that inhaled air pollution doses were lowest for electric bus passengers. Conclusions Commuters’ rush hour exposures were significantly influenced by mode of transport, route, and fuel type. PMID:20185385

  9. Evaluation of U-Zr hydride fuel for a thorium fuel cycle in an RTR concept

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Taek; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1999-12-31

    In this paper, we performed a design study of a thorium fueled reactor according to the design concept of the Radkowsky Thorium Reactor (RTR) and evaluated its overall performance. To enhance its performance and alleviate its problems, we introduced a new metallic uranium fuel, uranium-zirconium hydride (U-ZrH{sub 1.6}), as a seed fuel. For comparison, typical ABB/CE-type PWR based on SYSTEM 80+and standard RTR-type thorium reactor were also studied. From the results of performance analysis, we could ascertain advantages of RTR-type thorium fueled reactor in proliferation resistance, fuel cycle economics, and back-end fuel cycle. Also, we found that enhancement of proliferation resistance and safer operating conditions may be achieved by using the U-ZrH{sub 1.6} fuel in the seed region without additional penalties in comparison with the standard RTR`s U-Zr fuel. 6 refs., 2 figs., 6 tabs. (Author)

  10. Evaluation of U-Zr hydride fuel for a thorium fuel cycle in an RTR concept

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Taek; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    In this paper, we performed a design study of a thorium fueled reactor according to the design concept of the Radkowsky Thorium Reactor (RTR) and evaluated its overall performance. To enhance its performance and alleviate its problems, we introduced a new metallic uranium fuel, uranium-zirconium hydride (U-ZrH{sub 1.6}), as a seed fuel. For comparison, typical ABB/CE-type PWR based on SYSTEM 80+and standard RTR-type thorium reactor were also studied. From the results of performance analysis, we could ascertain advantages of RTR-type thorium fueled reactor in proliferation resistance, fuel cycle economics, and back-end fuel cycle. Also, we found that enhancement of proliferation resistance and safer operating conditions may be achieved by using the U-ZrH{sub 1.6} fuel in the seed region without additional penalties in comparison with the standard RTR`s U-Zr fuel. 6 refs., 2 figs., 6 tabs. (Author)

  11. Discrimination of source reactor type by multivariate statistical analysis of uranium and plutonium isotopic concentrations in unknown irradiated nuclear fuel material.

    Science.gov (United States)

    Robel, Martin; Kristo, Michael J

    2008-11-01

    The problem of identifying the provenance of unknown nuclear material in the environment by multivariate statistical analysis of its uranium and/or plutonium isotopic composition is considered. Such material can be introduced into the environment as a result of nuclear accidents, inadvertent processing losses, illegal dumping of waste, or deliberate trafficking in nuclear materials. Various combinations of reactor type and fuel composition were analyzed using Principal Components Analysis (PCA) and Partial Least Squares Discriminant Analysis (PLSDA) of the concentrations of nine U and Pu isotopes in fuel as a function of burnup. Real-world variation in the concentrations of (234)U and (236)U in the fresh (unirradiated) fuel was incorporated. The U and Pu were also analyzed separately, with results that suggest that, even after reprocessing or environmental fractionation, Pu isotopes can be used to determine both the source reactor type and the initial fuel composition with good discrimination.

  12. Computational simulation of the microstructure of irradiation damaged regions for the plate type fuel of UO2 microspheres dispersed in stainless steel matrix

    International Nuclear Information System (INIS)

    Reis, S.C. dos; Lage, A.F.; Braga, D.; Ferraz, W.B.

    2006-01-01

    Plate type fuel elements have high efficiency of thermal transference what benefits the heat flux with high rates of power output. In reactor cores, fuel elements, in general, are subject to a high neutrons flux, high working temperatures, severe corrosion conditions, direct interference of fission products that result from nuclear reactions and radiation interaction-matter. For plate type fuels composed of ceramic particles dispersed in metallic matrix, one can observe the damage regions that arise due to the interaction fission products in the metallic matrix. Aiming at evaluating the extension of the damage regions in function of the particles and its diameters, in this paper, computational geometric simulations structure of plate type fuel cores, composed of UO 2 microspheres dispersed in stainless steel in several fractions of volume and diameters were carried out. The results of the simulations were exported to AutoCAD R where it was possible its visualization and analysis. (author)

  13. Bilateral Vestibular Dysfunction Associated With Chronic Exposure to Military Jet Propellant Type-Eight Jet Fuel

    Directory of Open Access Journals (Sweden)

    Terry D. Fife

    2018-05-01

    Full Text Available We describe three patients diagnosed with bilateral vestibular dysfunction associated with the jet propellant type-eight (JP-8 fuel exposure. Chronic exposure to aromatic and aliphatic hydrocarbons, which are the main constituents of JP-8 military aircraft jet fuel, occurred over 3–5 years’ duration while working on or near the flight line. Exposure to toxic hydrocarbons was substantiated by the presence of JP-8 metabolite n-hexane in the blood of one of the cases. The presenting symptoms were dizziness, headache, fatigue, and imbalance. Rotational chair testing confirmed bilateral vestibular dysfunction in all the three patients. Vestibular function improved over time once the exposure was removed. Bilateral vestibular dysfunction has been associated with hydrocarbon exposure in humans, but only recently has emphasis been placed specifically on the detrimental effects of JP-8 jet fuel and its numerous hydrocarbon constituents. Data are limited on the mechanism of JP-8-induced vestibular dysfunction or ototoxicity. Early recognition of JP-8 toxicity risk, cessation of exposure, and customized vestibular therapy offer the best chance for improved balance. Bilateral vestibular impairment is under-recognized in those chronically exposed to all forms of jet fuel.

  14. Closed-form solution of a two-dimensional fuel temperature model for TRIGA-type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rivard, J B [Sandia Laboratories (United States)

    1974-07-01

    If azimuthal power density variations are ignored, the steady-state temperature distribution within a TRIGA-type fuel element is given by the solution of the Poisson equation in two dimensions (r and z) . This paper presents a closed-form solution of this equation as a function of the axial and radial power density profiles, the conductivity of the U-ZrH, the inlet temperature, specific heat and flow rate of the coolant, and the overall heat transfer coefficient. The method begins with the development of a system of linear ordinary differential equations describing mass and energy balances in the fuel and coolant. From the solution of this system, an expression for the second derivative of the fuel temperature distribution in the axial (z) direction is found. Substitution of this expression into the Poisson equation for T(r,z) reduces it from a partial differential equation to an ordinary differential equation in r, which is subsequently solved in closed-form. The results of typical calculations using the model are presented. (author)

  15. Performance analysis of hybrid solid oxide fuel cell and gas turbine cycle: Application of alternative fuels

    International Nuclear Information System (INIS)

    Zabihian, Farshid; Fung, Alan S.

    2013-01-01

    Highlights: • Variation of the stream properties in the syngas-fueled hybrid SOFC–GT cycle. • Detailed analysis of the operation of the methane-fueled SOFC–GT cycle. • Investigate effects of inlet fuel type and composition on performance of cycle. • Comparison of system operation when operated with and without anode recirculation. - Abstract: In this paper, the hybrid solid oxide fuel cell (SOFC) and gas turbine (GT) model was applied to investigate the effects of the inlet fuel type and composition on the performance of the cycle. This type of analysis is vital for the real world utilization of manufactured fuels in the hybrid SOFC–GT system due to the fact that these fuel compositions depends on the type of material that is processed, the fuel production process, and process control parameters. In the first part of this paper, it is shown that the results of a limited number of studies on the utilization of non-conventional fuels have been published in the open literature. However, further studies are required in this area to investigate all aspects of the issue for different configurations and assumptions. Then, the results of the simulation of the syngas-fueled hybrid SOFC–GT cycle are employed to explain the variation of the stream properties throughout the cycle. This analysis can be very helpful in understanding cycle internal working and can provide some interesting insights to the system operation. Then, the detailed information of the operation of the methane-fueled SOFC–GT cycle is presented. For both syngas- and methane-fueled cycles, the operating conditions of the equipment are presented and compared. Moreover, the comparison of the characteristics of the system when it is operated with two different schemes to provide the required steam for the cycle, with anode recirculation and with an external source of water, provides some interesting insights to the system operation. For instance, it was shown that although the physical

  16. Fuel property effects on Navy aircraft fuel systems

    Science.gov (United States)

    Moses, C. A.

    1984-01-01

    Problems of ensuring compatibility of Navy aircraft with fuels that may be different than the fuels for which the equipment was designed and qualified are discussed. To avoid expensive requalification of all the engines and airframe fuel systems, methodologies to qualify future fuels by using bench-scale and component testing are being sought. Fuel blends with increasing JP5-type aromatic concentration were seen to produce less volume swell than an equivalent aromatic concentration in the reference fuel. Futhermore, blends with naphthenes, decalin, tetralin, and naphthalenes do not deviate significantly from the correlation line of aromatic blends, Similar results are found with tensile strenth and elongation. Other elastomers, sealants, and adhesives are also being tested.

  17. Nuclear fuel string assembly

    International Nuclear Information System (INIS)

    Ip, A.K.; Koyanagi, K.; Tarasuk, W.R.

    1976-01-01

    A method of fabricating rodded fuels suitable for use in pressure tube type reactors and in pressure vessel type reactors is described. Fuel rods are secured as an inner and an outer sub-assembly, each rod attached between mounting rings secured to the rod ends. The two sub-assemblies are telescoped together and positioned by spaced thimbles located between them to provide precise positioning while permittng differential axial movement between the sub-assemblies. Such sub-assemblies are particularly suited for mounting as bundle strings. The method provides particular advantages in the assembly of annular-section fuel pins, which includes booster fuel containing enriched fuel material. (LL)

  18. Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

    International Nuclear Information System (INIS)

    Boer, Brian

    2007-01-01

    The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

  19. Nuclear reactor using fuel sphere for combustion and fuel spheres for breeding

    International Nuclear Information System (INIS)

    Yamashita, Kiyonobu.

    1995-01-01

    The present invention concerns a pebble bed-type reactor which can efficiently convert parent nuclides to fission nuclides. Fuel spheres for combustion having fission nuclides as main fuels, and fuel spheres for breeding having parent nuclides as main fuels are used separately, in the pebble bed-type reactor. According to the present invention, fuel spheres for breeding can be stayed in a reactor core for a long period of time, so that parent nuclides can be sufficiently converted into fission nuclides. In addition, since fuel spheres for breeding are loaded repeatedly, the amount thereof to be used is reduced. Therefore, the amount of the fuel spheres for breeding is small even when they are re-processed. On the other hand, since the content of the fission nuclides in the fuel spheres for breeding is not great, they can be put to final storage. This is attributable that although the fuel spheres for breeding contain fission nuclides generated by conversion, the fission nuclides are annihilated by nuclear fission reactions at the same time with the generation thereof. (I.S.)

  20. Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

    Directory of Open Access Journals (Sweden)

    Itamar Iliuk

    2016-01-01

    Full Text Available Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was proposed. A partial loss of flow accident was simulated to show the behavior of the model under this type of accident. The results show that the critical heat flux was detected in the central region along the axial direction of the plate when the right water channel was blocked.

  1. Method of detecting fuel failure in FBR type reactor and method of estimating fuel failure position

    International Nuclear Information System (INIS)

    Sonoda, Yukio; Tamaoki, Tetsuo

    1989-01-01

    Noise components in a normal state contained in detection signals from delayed neutron monitors disposed to a coolant inlet, etc. of an intermediate heat exchanger are forecast by self-recurring model and eliminated, and resultant detection signals are monitored thereby detecting fuel failure high sensitivity. Subsequently, the reactor is controlled to a low power operation state and a new self-recurring model to the detection signals from the delayed neutron monitors are prepared. Then, noise components in this state are removed and control rods near the delayed neutron monitors are extracted in a short stroke successively to examine the change of response of the delayed neutron monitors. Accordingly, the failed position for each of the fuels can be estimated at a level of one fuel assembly or a level of several assemblies containing the above-mentioned fuel assembly. Since the fuel failure can be detected at a high sensitivity and the position can be estimated, diffusion of abnormality can be prevented and plant shutdown for fuel exchange can be minimized. (I.S.)

  2. Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.; Billone, M.C.; Kramer, J.M.

    1992-01-01

    The next step in the development of metal fuels for the integral fast reactor (IFR) is the conversion of the Experimental Breeder Reactor II (EBR-II) core to one containing the ternary U-20 Pu-10 Zr alloy clad with HT-9 cladding, i.e., the Mk-V core. This paper presents results of three hot-cell furnace simulation tests on irradiated Mk-V-type fuel elements (U-19 Pu-10 Zr/HT-9), which were performed to support the safety case for the Mk-V core. These tests were designed to envelop an umbrella (bounding) unlikely loss-of-flow (LOF) event in EBR-II during which the calculated peak cladding temperature would reach 776 degree C for < 2 min. The principal objectives of these tests were (a) demonstration of the safety margin of the fuel element, (b) investigation of cladding breaching behavior, and (c) provision of data for validation of the FPIN2 and LIFE-METAL codes

  3. The relevance of the IFPE Database to the modelling of WWER-type fuel behaviour

    International Nuclear Information System (INIS)

    Killeen, J.; Sartori, E.

    2006-01-01

    The aim of the International Fuel Performance Experimental Database (IFPE Database) is to provide, in the public domain, a comprehensive and well-qualified database on zircaloy-clad UO 2 fuel for model development and code validation. The data encompass both normal and off-normal operation and include prototypic commercial irradiations as well as experiments performed in Material Testing Reactors. To date, the Database contains over 800 individual cases, providing data on fuel centreline temperatures, dimensional changes and FGR either from in-pile pressure measurements or PIE techniques, including puncturing, Electron Probe Micro Analysis (EPMA) and X-ray Fluorescence (XRF) measurements. This work in assembling and disseminating the Database is carried out in close co-operation and co-ordination between OECD/NEA and the IAEA. The majority of data sets are dedicated to fuel behaviour under LWR irradiation, and every effort has been made to obtain data representative of BWR, PWR and WWER conditions. In each case, the data set contains information on the pre-characterisation of the fuel, cladding and fuel rod geometry, the irradiation history presented in as much detail as the source documents allow, and finally any in-pile or PIE measurements that were made. The purpose of this paper is to highlight data that are relevant specifically to WWER application. To this end, the NEA and IAEA have been successful in obtaining appropriate data for both WWER-440 and WWER-1000-type reactors. These are: 1) Twelve (12) rods from the Finnish-Russian co-operative SOFIT programme; 2) Kola-3 WWER-440 irradiation; 3) MIR ramp tests on Kola-3 rods; 4) Zaporozskaya WWER-1000 irradiation; 5) Novovoronezh WWER-1000 irradiation. Before reviewing these data sets and their usefulness, the paper touches briefly on recent, more novel additions to the Database and on progress made in the use of the Database for the current IAEA FUMEX II Project. Finally, the paper describes the Computer

  4. Fuel assembly

    International Nuclear Information System (INIS)

    Chaki, Masao; Nishida, Koji; Karasawa, Hidetoshi; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Kashiwai, Shin-ichi; Masuhara, Yasuhiro

    1998-01-01

    The present invention concerns a fuel assembly, for a BWR type nuclear reactor, comprising fuel rods in 9 x 9 matrix. The inner width of the channel box is about 132mm and the length of the fuel rods which are not short fuel rods is about 4m. Two water rods having a circular cross section are arranged on a diagonal line in a portion of 3 x 3 matrix at the center of the fuel assembly, and two fuel rods are disposed at vacant spaces, and the number of fuel rods is 74. Eight fuel rods are determined as short fuel rods among 74 fuel rods. Assuming the fuel inventory in the short fuel rod as X(kg), and the fuel inventory in the fuel rods other than the short fuel rods as Y(kg), X and Y satisfy the relation: X + Y ≥ 173m, Y ≤ - 9.7X + 292, Y ≤ - 0.3X + 203 and X > 0. Then, even when the short fuel rods are used, the fuel inventory is increased and fuel economy can be improved. (I.N.)

  5. Intelligent support system online for the operation of fossil fuel units; Sistema inteligente de ayuda en linea para la operacion de unidades termoelectricas

    Energy Technology Data Exchange (ETDEWEB)

    Quintero R, Agustin; Suarez C, Dionisio A; Sanchez L, Jose Alfredo [Instituto de Investigaciones Electricas, Cuernavaca, Morelos (Mexico)

    2006-07-01

    The purpose of this publication is to present a support system online for the operation (SAO) of fossil fuel units, having as objective to support the operator of fossil fuel units when handling guidelines and essential information to carry out the starting and shutdown in a safe and efficient way, reducing the consumption of the useful life of the steam generator and the turbine, as well as the operational cost. Here are presented the intelligent systems for the support of the operation and are described the main characteristics in the dominion of application for the support system, its architecture and functionality, as well as the results obtained in the system assessment performed with the aid of total reach simulator of a fossil fuel unit and a pilot unit. [Spanish] Es motivo de esta publicacion, la presentacion de un sistema de ayuda en linea para la operacion (SAO) de unidades termoelectricas que tiene como objetivo asistir al operador de una unidad termoelectrica con guias de manejo e informacion esencial para llevar a cabo el arranque o paro en forma segura y eficiente, reducir el consumo de vida util del generador de vapor y la turbina, asi como el costo de operacion. Aqui se presentan los sistemas inteligentes de ayuda a la operacion y se describen las caracteristicas principales del dominio de aplicacion para el sistema de ayuda, su arquitectura y funcionalidad, asi como los resultados obtenidos de la evaluacion del sistema realizada con la ayuda de un simulador de alcance total de una unidad termoelectrica y una unidad piloto.

  6. Unified fuel elements development for research reactors

    International Nuclear Information System (INIS)

    Vatulin, A.; Stetsky, Y.; Dobrikova, I.

    1998-01-01

    Square cross-section rod type fuel elements have been developed for russian pool-type research reactors. new fuel elements can replace the large nomenclature of tubular fuel elements with around, square and hexahedral cross-sections and to solve a problem of enrichment reduction. the fuel assembly designs with rod type fuel elements have been developed. The overall dimensions of existing the assemblies are preserved in this one. the experimental-industrial fabricating process of fuel elements, based on a joint extrusion method has been developed. The fabricating process has been tested in laboratory conditions, 150 experimental fuel element samples of the various sizes were produced. (author)

  7. Hybrid two fuel system nozzle with a bypass connecting the two fuel systems

    Science.gov (United States)

    Varatharajan, Balachandar [Cincinnati, OH; Ziminsky, Willy Steve [Simpsonville, SC; Yilmaz, Ertan [Albany, NY; Lacy, Benjamin [Greer, SC; Zuo, Baifang [Simpsonville, SC; York, William David [Greer, SC

    2012-05-29

    A hybrid fuel combustion nozzle for use with natural gas, syngas, or other types of fuels. The hybrid fuel combustion nozzle may include a natural gas system with a number of swozzle vanes and a syngas system with a number of co-annular fuel tubes.

  8. Gas chromatography-mass spectrometric determination of traces of ether-type icing inhibitors in free-floating fuels

    Energy Technology Data Exchange (ETDEWEB)

    Shin, H.S. [Dept. of Environmental Education, Kongju National Univ., Kongju (Korea); Abuse Drug Research Center, Kongju National Univ., Kongju (Korea); Ahn, H.S. [Dept. of Environmental Science, Kongju National Univ., Kongju (Korea)

    2004-08-01

    A gas chromatographic-mass spectrometric (GC-MS) assay method has been developed for simultaneous determination of ethylene glycol monomethyl ether (EGME) and diethylene glycol monomethly ether (DEGME) in spilled aviation fuels. Ethylene glycol monobutyl ether (EGBE) and ethylene glycol monoethyl ether (EGEE) were used as internal standard and surrogate, respectively. Sample preparation consisted of back-extraction with 7 mL dichloromethane after extraction of 50 mL of fuel with 2 mL of water. The extract was concentrated to dryness, dissolved in 100 {mu}L methanol, and analyzed by GC-MS with selected-ion monitoring (SIM). The peaks had good chromatographic properties on a semi-polar column. EGME and DEGME were extracted from fuel with high recovery of 75 and 85%, with small variations, respectively. Method detection limits were 1.3 and 1.0 ng mL{sup -1} for EGME and DEGME, respectively, in spilled fuel. DEGME was detected at concentrations of 22.6 and 19.7 ng mL{sup -1} in two samples from among five free-floating samples collected in a tunnel of a subway station located in the vicinity of an army base in Korea. The method might be useful for differentiation between the fuel-types kerosene and JP-8, which might originate from a storage tank. (orig.)

  9. Advanced technique for computing fuel combustion properties in pulverized-fuel fired boilers

    Energy Technology Data Exchange (ETDEWEB)

    Kotler, V.R. (Vsesoyuznyi Teplotekhnicheskii Institut (Russian Federation))

    1992-03-01

    Reviews foreign technical reports on advanced techniques for computing fuel combustion properties in pulverized-fuel fired boilers and analyzes a technique developed by Combustion Engineering, Inc. (USA). Characteristics of 25 fuel types, including 19 grades of coal, are listed along with a diagram of an installation with a drop tube furnace. Characteristics include burn-out intensity curves obtained using thermogravimetric analysis for high-volatile bituminous, semi-bituminous and coking coal. The patented LFP-SKM mathematical model is used to model combustion of a particular fuel under given conditions. The model allows for fuel particle size, air surplus, load, flame height, and portion of air supplied as tertiary blast. Good agreement between computational and experimental data was observed. The method is employed in designing new boilers as well as converting operating boilers to alternative types of fuel. 3 refs.

  10. An optimized BWR fuel lattice for improved fuel utilization

    International Nuclear Information System (INIS)

    Bernander, O.; Helmersson, S.; Schoen, C.G.

    1984-01-01

    Optimization of the BWR fuel lattice has evolved into the water cross concept, termed ''SVEA'', whereby the improved moderation within bundles augments reactivity and thus improves fuel cycle economy. The novel design introduces into the assembly a cruciform and double-walled partition containing nonboiling water, thus forming four subchannels, each of which holds a 4x4 fuel rod bundle. In Scandinavian BWRs - for which commercial SVEA reloads are now scheduled - the reactivity gain is well exploited without adverse impact in other respects. In effect, the water cross design improves both mechanical and thermal-hydraulic performance. Increased average burnup is also promoted through achieving flatter local power distributions. The fuel utilization savings are in the order of 10%, depending on the basis of comparison, e.g. choice of discharge burnup and lattice type. This paper reviews the design considerations and the fuel utilization benefits of the water cross fuel for non-Scandinavian BWRs which have somewhat different core design parameters relative to ASEA-ATOM reactors. For one design proposal, comparisons are made with current standard 8x8 fuel rod bundles as well as with 9x9 type fuel in reactors with symmetric or asymmetric inter-assembly water gaps. The effect on reactivity coefficients and shutdown margin are estimated and an assessment is made of thermal-hydraulic properties. Consideration is also given to a novel and advantageous way of including mixed-oxide fuel in BWR reloads. (author)

  11. Adsorptive on-board desulfurization over multiple cycles for fuel-cell-based auxiliary power units operated by different types of fuels

    Science.gov (United States)

    Neubauer, Raphael; Weinlaender, Christof; Kienzl, Norbert; Bitschnau, Brigitte; Schroettner, Hartmuth; Hochenauer, Christoph

    2018-05-01

    On-board desulfurization is essential to operate fuel-cell-based auxiliary power units (APU) with commercial fuels. In this work, both (i) on-board desulfurization and (ii) on-board regeneration performance of Ag-Al2O3 adsorbent is investigated in a comprehensive manner. The herein investigated regeneration strategy uses hot APU off-gas as the regeneration medium and requires no additional reagents, tanks, nor heat exchangers and thus has remarkable advantages in comparison to state-of-the-art regeneration strategies. The results for (i) show high desulfurization performance of Ag-Al2O3 under all relevant operating conditions and specify the influence of individual operation parameters and the combination of them, which have not yet been quantified. The system integrated regeneration strategy (ii) shows excellent regeneration performance recovering 100% of the initial adsorption capacity for all investigated types of fuels and sulfur heterocycles. Even the adsorption capacity of the most challenging dibenzothiophene in terms of regeneration is restored to 100% over 14 cycles of operation. Subsequent material analyses proved the thermal and chemical stability of all relevant adsorption sites under APU off-gas conditions. To the best of our knowledge, this is the first time 100% regeneration after adsorption of dibenzothiophene is reported over 14 cycles of operation for thermal regeneration in oxidizing atmospheres.

  12. Fuel cells 101

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, B.

    2003-06-01

    A capsule history of fuel cells is given, beginning with the first discovery in 1839 by William Grove, a Welsh judge who, when experimenting with electrolysis discovered that by re-combining the two components of electrolysis (water and oxygen) an electric charge was produced. A century later, in 1958, Francis Thomas Bacon, a British scientist demonstrated the first working fuel cell stack, a technology which was licensed and used in the Apollo spacecraft. In Canada, early research on the development of fuel cells was carried out at the University of Toronto, the Defence Research Establishment and the National Research Council. Most of the early work concentrated on alkaline and phosphoric acid fuel cells. In 1983, Ballard Research began the development of the electrolyte membrane fuel cell, which marked the beginning of Canada becoming a world leader in fuel cell technology development. The paper provides a brief account of how fuel cells work, describes the distinguishing characteristics of the various types of fuel cells (alkaline, phosphoric acid, molten-carbonate, solid oxide, and proton exchange membrane types) and their principal benefits. The emphasis is on proton exchange membrane fuel cells because they are the only fuel cell technology that is appropriate for providing primary propulsion power onboard a vehicle. Since vehicles are by far the greatest consumers of fossil fuels, it follows that proton exchange membrane fuel cells will have the greatest potential impact on both environmental matters and on our reliance on oil as our primary fuel. Various on-going and planned fuel cell demonstration projects are also described. 1 fig.

  13. Plant Performance of Solid Oxide Fuel Cell Systems Fed by Alternative Fuels

    DEFF Research Database (Denmark)

    Rokni, Masoud

    2016-01-01

    Different plant design for several fuel types such as natural gas, methanol, ethanol, DME, ammonia and pure hydrogen are presented and analysed. Anode recirculation which is an important issue in SOFC plants are also explored and studied. It is shown that depending on type of the fuel whether fuel...... recycle increases plant efficiency only if fuel utilization factor is low. Other important issues such as why plant efficiency is lower when it is fed with hydrogen or biogas compared to when it is fed by other fuels such as methanol, ethanol, DME and ammonia will also be discussed and explained....... For example, plant efficiency of 45%, 54% and 50.5% can be achieved if the hydrogen, ethanol and methanol are used respectively....

  14. Thermal breeder fuel enrichment zoning

    International Nuclear Information System (INIS)

    Capossela, H.J.; Dwyer, J.R.; Luce, R.G.; McCoy, D.F.; Merriman, F.C.

    1992-01-01

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect. 1 figure

  15. Assessment of fuel damage of pool type research reactor in the case of fuel plates blockage

    Energy Technology Data Exchange (ETDEWEB)

    Jalil, Jafari; Samad, Khakshournia [AEOI, Karegar Ave. School of R and D of Nuclear Reactors and Accelerators, Teheran (Iran, Islamic Republic of); D' Auria, F. [Pisa Univ., DIMNP (Italy)

    2007-07-01

    Tehran Research Reactor (TRR) is a pool type 5 MW research reactor. It is assumed that external objects or debris that may fall down to reactor core cause obstruction of coolant flow through one of the fuel assemblies. Thermal hydraulic analysis of this event, using the RELAP5 system code has been studied. The reported transient is related to the partial and total obstruction of a single Fuel Element (FE) cooling channel of 27 FE equilibrium core of TRR. Such event constitutes a severe accident for this type of reactor since it may lead to local dryout and eventually to loss of the FE integrity. Two scenarios are analysed to emphasize the severity of the accident. The first one is a partial blockage of an average FE considering four different obstruction levels: 25%, 50%, 75% and 97% of nominal flow area. The second one is an extreme scenario consisting of total blockage of the same FE. This study constitutes the first step of a larger work which consists of performing a 3-dimensional simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic calculation included in the RELAP5 code. Main results obtained from the RELAP5 calculations are as following. First, in the case of flow blockage under 97% of the nominal flow area of an average FE, only an increase of the coolant and clad temperatures is observed without any consequences for the integrity of the FE. The mass flow rate remains sufficient to cool the clad safely. Secondly, in the case of total obstruction of the nominal flow area, it is seen that transient turns out to be a severe accident due to the dryout conditions are reached shortly and melting of the cladding occurs. Thirdly, the use of the point kinetic approach leads to conservative results. A best estimate simulation of such kind of transients requires the use of 3-dimensional kinetic calculations, which could be done using the current Coupled Codes

  16. Fuel performance in water storage

    International Nuclear Information System (INIS)

    Hoskins, A.P.; Scott, J.G.; Shelton-Davis, C.V.; McDannel, G.E.

    1993-11-01

    Westinghouse Idaho Nuclear Company operates the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory (INEL) for the Department of Energy (DOE). A variety of different types of fuels have been stored there since the 1950's prior to reprocessing for uranium recovery. In April of 1992, the DOE decided to end fuel reprocessing, changing the mission at ICPP. Fuel integrity in storage is now viewed as long term until final disposition is defined and implemented. Thus, the condition of fuel and storage equipment is being closely monitored and evaluated to ensure continued safe storage. There are four main areas of fuel storage at ICPP: an original underwater storage facility (CPP-603), a modern underwater storage facility (CPP-666), and two dry fuel storage facilities. The fuels in storage are from the US Navy, DOE (and its predecessors the Energy Research and Development Administration and the Atomic Energy Commission), and other research programs. Fuel matrices include uranium oxide, hydride, carbide, metal, and alloy fuels. In the underwater storage basins, fuels are clad with stainless steel, zirconium, and aluminum. Also included in the basin inventory is canned scrap material. The dry fuel storage contains primarily graphite and aluminum type fuels. A total of 55 different fuel types are currently stored at the Idaho Chemical Processing Plant. The corrosion resistance of the barrier material is of primary concern in evaluating the integrity of the fuel in long term water storage. The barrier material is either the fuel cladding (if not canned) or the can material

  17. Study of core characteristics on fuel and coolant type. Results of F/S phase-I

    International Nuclear Information System (INIS)

    Ikegami, Tetsuo; Hayashi, Hideyuki; Sasaki, Makoto; Mizuno, Tomoyasu; Yamadate, Megumi; Takaki, Naoyuki; Kurosawa, Norifumi; Sakashita, Yoshiaki; Naganuma, Masayuki

    2001-03-01

    The phase-I of the Feasibility Study of Commercialized Fast Reactor Cycle Systems (F/S) were started from July, 1999 and terminated at the end of FY2000 in order to executed examination about technology alternatives of various commercialized fast reactor (FR) recycle concepts, in response to the JNC middle long term enterprise plan. In the phase-I of this F/S, a number of conceptual candidates have been selected from the following 5 viewpoints: a) ensuring safety, b) economic competitiveness to future LWRs, c) efficient utilization of resources, d) reduction of environmental burden, e) enhancement of nuclear non-proliferation. As for this study from the above viewpoints, core characteristics of many kinds of reactors have been investigated, analyzed and examined a core / a fuel characteristic in the combinations of fuel and coolant types and power output scales. Based on these results, R and D plans of the phase-II to be performed have been proposed, and a database to select candidate reactor concepts has been prepared. The conclusions have been obtained in the phase-I are as follows: (1) Evaluation of a fuel form in every each coolant was compared. A promising fuel form was extracted as follows: an oxide and a metal fuel for sodium coolant cores, a metal and a nitride fuel for heavy metal coolant cores, an oxide and a nitride fuel for carbon dioxide coolant cores and a nitride fuel for He gas coolant cores. (2) As the general idea that performance of a core nucleus can be compatible with re-criticality evasion in sodium coolant large-sized oxide fuel cores, a axial blanket particle elimination radial heterogeneous core is one influential candidate. (3) In case of Pb-Bi coolant nature circulation medium size core with an oxide fuel, it is difficult to simultaneously achieve higher discharged burn-up and higher breeding ratio according to the viewpoints of the phase-I. (4) Core characteristics of a carbon dioxide coolant core shows to be almost equivalent to that of

  18. A comparison of sodium borohydride as a fuel for proton exchange membrane fuel cells and for direct borohydride fuel cells

    Science.gov (United States)

    Wee, Jung-Ho

    Two types of fuel cell systems using NaBH 4 aqueous solution as a fuel are possible: the hydrogen/air proton exchange membrane fuel cell (PEMFC) which uses onsite H 2 generated via the NaBH 4 hydrolysis reaction (B-PEMFC) at the anode and the direct borohydride fuel cell (DBFC) system which directly uses NaBH 4 aqueous solution at the anode and air at the cathode. Recently, research on these two types of fuel cells has begun to attract interest due to the various benefits of this liquid fuel for fuel cell systems for portable applications. It might therefore be relevant at this stage to evaluate the relative competitiveness of the two fuel cells. Considering their current technologies and the high price of NaBH 4, this paper evaluated and analyzed the factors influencing the relative favorability of each type of fuel cell. Their relative competitiveness was strongly dependent on the extent of the NaBH 4 crossover. When considering the crossover in DBFC systems, the total costs of the B-PEMFC system were the most competitive among the fuel cell systems. On the other hand, if the crossover problem were to be completely overcome, the total cost of the DBFC system generating six electrons (6e-DBFC) would be very similar to that of the B-PEMFC system. The DBFC system generating eight electrons (8e-DBFC) became even more competitive if the problem of crossover can be overcome. However, in this case, the volume of NaBH 4 aqueous solution consumed by the DBFC was larger than that consumed by the B-PEMFC.

  19. A Study on BC Emission from Vehicles using Different Types of Fuel

    Science.gov (United States)

    Kim, K.; Son, J.; Kim, J.; Kim, S.; Park, G.; Sung, K.; Kim, I.; Chung, T.; Park, T.; Kang, S.; Ban, J.; Kim, J.; Hong, Y. D.; Woo, J. H.; Lee, T.

    2017-12-01

    Black carbon (BC) is an anthropogenic aerosol from fossil fuels, and biomass burning. It absorbs solar radiation, and heats the atmosphere leading 0.4W m-2 radiative forcing. BC is a particle that can cause serious effects on human body as well. Toxicological studies of black carbon suggests that BC may be an important carrier of toxic chemicals to human body. The recent researches show that one of the main precursor of BC is vehicle emission, but the inventory of BC emission rate from vehicle is inadequate in South Korea. This study tries to find differences of BC emission from different sizes of vehicles using different types of fuels. Fuels used in vehicles are gasoline, liquefied petroleum gas (LPG), and diesel. BC was directly measured from the tail pipe of vehicles using Aethalometer (AE33, Magee Scientific Corporation). This study was conducted in Transport Pollutant Research Center, National Institute of Environmental Research, South Korea. Measurement was progressed with the five different test modes of speeds. Speed modes includes 4.7, 17.3, 34.1, 65.4, and 97.3 km h-1. Emission rate of BC was high in the slowest speed mode, and showed decrease with increase of the speed of vehicles. Gasoline vehicles had the relatively higher emission rate of BC than the LPG vehicle, while the emission rate of BC for Diesel with DPF (Diesel Particle Filter) was observed to be the lowest.

  20. Analysis of gamma heating at TRIGA mark reactor core Bandung using plate type fuel

    International Nuclear Information System (INIS)

    Setiyanto; Tukiran Surbakti

    2016-01-01

    In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities) and central irradiation position (CIP), especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0.87 W/g), but very low value for Lazy Susan position (lest then 0.11 W/g). Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. (author)

  1. Power distribution investigation in the transition phase of the low moderation type MOX fueled LWR from the high conversion core to the breeding core

    International Nuclear Information System (INIS)

    Akie, Hiroshi; Nakano, Yoshihiro; Okubo, Tsutomu

    2011-01-01

    The key concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is a core transition from a high conversion (HC) type to a plutonium breeding (BR) type in a same reactor system only by replacing fuel assemblies. Consequently in this transition phase, there are two types of assemblies in the same core. Due to the differences of the two assembly types, region-wise soft to hard neutron spectra appears and result in a large power peaking. Therefore, power distribution of FLWR in the HC to BR transition phase was studied by performing assembly and core calculations. For the whole core calculation, a new 14-group energy structure is developed to better represent the power distribution obtained with the fine 107-group structure than the 9-group structure in the previous evaluations. Calculations on few assemblies geometries show large local power peakings can be effectively reduced by considering plutonium enrichment distribution in an assembly. In the whole core calculation, there is a power level mismatch between HC and BR assemblies, but overall power distribution flattening is possible by optimizing fuel assemblies loading. Although the fuel loading should be decided also taking into account the void coefficient, transition from HC to BR type FLWR seems feasible without difficulty. (author)

  2. In-pile irradiation of rock-like oxide fuels

    International Nuclear Information System (INIS)

    Nitani, N.; Kuramoto, K.; Yamashita, T.; Nakano, Y.; Akie, H.

    2001-01-01

    Five kinds of ROX fuels were prepared and irradiated using 20% enriched U instead of Pu. Non-destructive and destructive post-irradiation examinations were carried out. FP gas release rates of the particle-dispersed type fuels and homogeneously-blended type fuels were larger than that of the Yttria-stabilized zirconia containing UO 2 single phase fuel. From results of SEM and EPMA, decomposition of the spinel was observed. The decomposition of the spinel is probably avoided by lowering the irradiation temperature, less than 1700 K. The regions suffering the irradiation damage of the particle dispersed type fuels were less than those of the homogeneously-blended type fuels. (author)

  3. In-pile irradiation of rock-like oxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nitani, N.; Kuramoto, K.; Yamashita, T.; Nakano, Y.; Akie, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2001-07-01

    Five kinds of ROX fuels were prepared and irradiated using 20% enriched U instead of Pu. Non-destructive and destructive post-irradiation examinations were carried out. FP gas release rates of the particle-dispersed type fuels and homogeneously-blended type fuels were larger than that of the Yttria-stabilized zirconia containing UO{sub 2} single phase fuel. From results of SEM and EPMA, decomposition of the spinel was observed. The decomposition of the spinel is probably avoided by lowering the irradiation temperature, less than 1700 K. The regions suffering the irradiation damage of the particle dispersed type fuels were less than those of the homogeneously-blended type fuels. (author)

  4. Tools for LWR spent fuel characterization: Assembly classes and fuel designs

    International Nuclear Information System (INIS)

    Moore, R.S.; Notz, K.J.

    1991-01-01

    The Characteristics Data Base (CDB) is sponsored by the DOE's Office of Civilian Radioactive Waste Management (OCRWM). The CDB provides a single, comprehensive source of data pertaining to radioactive wastes that will or may require geologic disposal, including detailed data describing the physical, quantitative, and radiological characteristics of light-water reactor (LWR) spent fuel. In developing the CDB, tools for the classification of fuel assembly types have been developed. The assembly class scheme is particularly useful for size- and handling-based describes these tools and presents results of their applications in the areas of fuel assembly type identification, characterization of projected discharges, cask accommodation analyses, and defective fuel analyses. Suggestions for additional applications are also made. 7 refs., 1 fig., 2 tabs

  5. Quantitative determination of uranium distribution homogeneity in MTR fuel type plates

    International Nuclear Information System (INIS)

    Ferrufino, Felipe Bonito Jaldin

    2011-01-01

    IPEN/CNEN-SP produces the fuel to supply its nuclear research reactor IEA-R1. The fuel is assembled with fuel plates containing an U 3 Si 2 -Al composite meat. A good homogeneity in the uranium distribution inside the fuel plate meat is important from the standpoint of irradiation performance. Considering the lower power of reactor IEA-R1, the uranium distribution in the fuel plate has been evaluated only by visual inspection of radiographs. However, with the possibility of IPEN to manufacture the fuel for the new Brazilian Multipurpose Reactor (RMB), with higher power, it urges to develop a methodology to determine quantitatively the uranium distribution into the fuel. This paper presents a methodology based on X-ray attenuation, in order to quantify the uranium concentration distribution in the meat of the fuel plate by using optical densities in radiographs and comparison with standards. The results demonstrated the inapplicability of the method, considering the current specification for the fuel plates due to the high intrinsic error to the method. However, the study of the errors involved in the methodology, seeking to increase their accuracy and precision, can enable the application of the method to qualify the final product. (author)

  6. High conversion burner type reactor

    International Nuclear Information System (INIS)

    Higuchi, Shin-ichi; Kawashima, Masatoshi

    1987-01-01

    Purpose: To simply and easily dismantle and reassemble densified fuel assemblies taken out of a high conversion ratio area thereby improve the neutron and fuel economy. Constitution: The burner portion for the purpose of fuel combustion is divided into a first burner region in adjacent with the high conversion ratio area at the center of the reactor core, and a second burner region formed to the outer circumference thereof and two types of fuels are charged therein. Densified fuel assemblies charged in the high conversion ratio area are separatably formed as fuel assemblies for use in the two types of burners. In this way, dense fuel assembly is separated into two types of fuel assemblies for use in burner of different number and arranging density of fuel elements which can be directly charged to the burner portion and facilitate the dismantling and reassembling of the fuel assemblies. Further, since the two types of fuel assemblies are charged in the burner portion, utilization factor for the neutron fuels can be improved. (Kamimura, M.)

  7. Apparatus for loading fuel pellets in fuel rods

    International Nuclear Information System (INIS)

    Tedesco, R.J.

    1976-01-01

    An apparatus is disclosed for loading fuel pellets into fuel rods for a nuclear reactor including a base supporting a table having grooves therein for holding a multiplicity of pellets. Multiple fuel rods are placed in alignment with grooves in the pellet table and a guide member channels pellets from the table into the corresponding fuel rods. To effect movement of pellets inside the fuel rods without jamming, a number of electromechanical devices mounted on the base have arms connected to the lower surface of the fuel rod table which cyclically imparts a reciprocating arc motion to the table for moving the fuel pellets longitudinally of and inside the fuel rods. These electromechanical devices include a solenoid having a plunger therein connected to a leaf type spring, the arrangement being such that upon energization of the solenoid coil, the leaf spring moves the fuel rod table rearwardly and downwardly, and upon deenergization of the coil, the spring imparts an upward-forward movement to the table which results in physical displacement of fuel pellets in the fuel rods clamped to the table surface. 8 claims, 6 drawing figures

  8. Seventh Edition Fuel Cell Handbook

    Energy Technology Data Exchange (ETDEWEB)

    NETL

    2004-11-01

    Provides an overview of fuel cell technology and research projects. Discusses the basic workings of fuel cells and their system components, main fuel cell types, their characteristics, and their development status, as well as a discussion of potential fuel cell applications.

  9. Emissions deterioration for three alternative fuel vehicle types: Natural gas, ethanol, and methanol vehicles

    International Nuclear Information System (INIS)

    Winebrake, J.J.; Deaton, M.L.

    1997-01-01

    Although there have been several studies examining emissions from in-use alternative fuel vehicles (AFVs), little is known about the deterioration of these emissions over vehicle lifetimes and how this deterioration compares with deterioration from conventional vehicles (CVs). This paper analyzes emissions data from 70 AFVs and 70 CVs operating in the federal government fleet to determine whether AFV emissions deterioration differs significantly from CV emissions deterioration. The authors conduct the analysis on three alternative fuel types (natural gas, methanol, and ethanol) and on five pollutants (carbon monoxide, carbon dioxide, total hydrocarbons, non-methane hydrocarbons, and nitrogen oxides). They find that for most cases they studied, deterioration differences are not statistically significant; however, several exceptions suggest that air quality planners and regulators must further analyze AFV emissions deterioration in order to properly include these technologies into broader air quality management schemes

  10. Fuel cycle

    International Nuclear Information System (INIS)

    Bahm, W.

    1989-01-01

    The situation of the nuclear fuel cycle for LWR type reactors in France and in the Federal Republic of Germany was presented in 14 lectures with the aim to compare the state-of-the-art in both countries. In addition to the momentarily changing fuilds of fuel element development and fueling strategies, the situation of reprocessing, made interesting by some recent developmnts, was portrayed and differences in ultimate waste disposal elucidated. (orig.) [de

  11. The thermal performance of a loop-type heat pipe for passively removing residual heat from spent fuel pool

    Energy Technology Data Exchange (ETDEWEB)

    Xiong, Zhenqin [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Gu, Hanyang, E-mail: guhanyang@stu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Wang, Minglu [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Cheng, Ye [Shanghai Nuclear Engineering Research and Design Institute, Shanghai 200233 (China)

    2014-12-15

    Highlights: • Feasibility of applying loop-type heat pipes for SFP is studied. • The heat transfer rate of the heat pipes was tested. • The heat transfer coefficient was between 200 and 490 W/m{sup 2}/s. • The effect of the water temperature is dominant. • Three kinds of the filling ratio 27%, 21% and 14% are compared. - Abstract: Heat pipe is an efficient heat transfer device without electrically driven parts. Therefore large-scale loop type heat pipe systems have potential uses for passively removing heat from spent fuel pools and reactor cores under the accidental conditions to improve the safety of the nuclear power station. However, temperature difference between the hot water in the spent fuel pool and the ambient air which is the heat sink is small, in the range of 20–60 °C. To understand and predict the heat removal capacity of such a large scale loop type heat pipe in the situation similar to the accidental condition of the spent fuel pool (SFP) for the design purpose, a loop-type heat pipe with a very high and large evaporator has been fabricated and was tested using ammonia as the working fluid. The evaporator with inner diameter of 65 mm and length of 7.6 m is immersed in a hot water tube which simulate the spent fuel pool. The condenser of the loop-type heat pipe is cooled by the air. The tests were performed with the velocity of the hot water in the tube in the range of 0.7–2.1 × 10{sup −2} m/s, the hot water inlet temperature between 50 and 90 °C and the air velocity ranging from 0.5 m/s to 2.5 m/s. Three kinds of the ammonia volumetric filling ratio in the heat pipe were tested, i.e. 27%, 21% and 14%. It is found that the heat transfer rate was in the range of 1.5–14.9 kW, and the heat transfer coefficient of evaporator was between 200 and 490 W/m{sup 2}/s. It is feasible to use the large scale loop type heat pipe to passively remove the residual heat from SFP. Furthermore, the effect of air velocity, air temperature, water flow

  12. The thermal performance of a loop-type heat pipe for passively removing residual heat from spent fuel pool

    International Nuclear Information System (INIS)

    Xiong, Zhenqin; Gu, Hanyang; Wang, Minglu; Cheng, Ye

    2014-01-01

    Highlights: • Feasibility of applying loop-type heat pipes for SFP is studied. • The heat transfer rate of the heat pipes was tested. • The heat transfer coefficient was between 200 and 490 W/m 2 /s. • The effect of the water temperature is dominant. • Three kinds of the filling ratio 27%, 21% and 14% are compared. - Abstract: Heat pipe is an efficient heat transfer device without electrically driven parts. Therefore large-scale loop type heat pipe systems have potential uses for passively removing heat from spent fuel pools and reactor cores under the accidental conditions to improve the safety of the nuclear power station. However, temperature difference between the hot water in the spent fuel pool and the ambient air which is the heat sink is small, in the range of 20–60 °C. To understand and predict the heat removal capacity of such a large scale loop type heat pipe in the situation similar to the accidental condition of the spent fuel pool (SFP) for the design purpose, a loop-type heat pipe with a very high and large evaporator has been fabricated and was tested using ammonia as the working fluid. The evaporator with inner diameter of 65 mm and length of 7.6 m is immersed in a hot water tube which simulate the spent fuel pool. The condenser of the loop-type heat pipe is cooled by the air. The tests were performed with the velocity of the hot water in the tube in the range of 0.7–2.1 × 10 −2 m/s, the hot water inlet temperature between 50 and 90 °C and the air velocity ranging from 0.5 m/s to 2.5 m/s. Three kinds of the ammonia volumetric filling ratio in the heat pipe were tested, i.e. 27%, 21% and 14%. It is found that the heat transfer rate was in the range of 1.5–14.9 kW, and the heat transfer coefficient of evaporator was between 200 and 490 W/m 2 /s. It is feasible to use the large scale loop type heat pipe to passively remove the residual heat from SFP. Furthermore, the effect of air velocity, air temperature, water flow rate and

  13. Denatured fuel cycles

    International Nuclear Information System (INIS)

    Till, C.E.

    1979-01-01

    This paper traces the history of the denatured fuel concept and discusses the characteristics of fuel cycles based on the concept. The proliferation resistance of denatured fuel cycles, the reactor types they involve, and the limitations they place on energy generation potential are discussed. The paper concludes with some remarks on the outlook for such cycles

  14. Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

    Science.gov (United States)

    Jiang, Yijie; Wang, Qiming; Cui, Yi; Huo, Yongzhong; Ding, Shurong

    2011-06-01

    Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent

  15. Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Jiang Yijie; Wang Qiming; Cui Yi; Huo Yongzhong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Ding Shurong, E-mail: dsr1971@163.com [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)

    2011-06-15

    Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent

  16. Polyvalent fuel treatment facility (TCP): shearing and dissolution of used fuel at La Hague facility

    Energy Technology Data Exchange (ETDEWEB)

    Brueziere, J.; Tribout-Maurizi, A.; Durand, L.; Bertrand, N. [Recycling Business Unit, AREVA, 1 place de la coupole, 92084 Paris La defense Cedex (France)

    2013-07-01

    Although many used nuclear fuel types have already been recycled, recycling plants are generally optimized for Light Water Reactor (LWR) UO{sub x} fuel. Benefits of used fuel recycling are consequently restricted to those fuels, with only limited capacity for the others like LWR MOX, Fast Reactor (FR) MOX or Research and Test Reactor (RTR) fuel. In order to recycle diverse fuel types, an innovative and polyvalent shearing and dissolving cell is planned to be put in operation in about 10 years at AREVA's La Hague recycling plant. This installation, called TCP (French acronym for polyvalent fuel treatment) will benefit from AREVA's industrial feedback, while taking part in the next steps towards a fast reactor fuel cycle development using innovative treatment solutions. Feasibility studies and R/Development trials on dissolution and shearing are currently ongoing. This new installation will allow AREVA to propose new services to its customers, in particular in term of MOX fuel, Research Test Reactors fuel and Fast Reactor fuel treatment. (authors)

  17. Effects of vehicle type and fuel quality on real world toxic emissions from diesel vehicles

    Science.gov (United States)

    Nelson, Peter F.; Tibbett, Anne R.; Day, Stuart J.

    Diesel vehicles are an important source of emissions of air pollutants, particularly oxides of nitrogen (NO x), particulate matter (PM), and toxic compounds with potential health impacts including volatile organic compounds (VOCs) such as benzene and aldehydes, and polycyclic aromatic hydrocarbons (PAHs). Current developments in engine design and fuel quality are expected to reduce these emissions in the future, but many vehicles exceed 10 years of age and may make a major contribution to urban pollutant concentrations and related health impacts for many years. In this study, emissions of a range of toxic compounds are reported using in-service vehicles which were tested using urban driving cycles developed for Australian conditions. Twelve vehicles were chosen from six vehicle weight classes and, in addition, two of these vehicles were driven through the urban drive cycle using a range of diesel fuel formulations. The fuels ranged in sulphur content from 24 to 1700 ppm, and in total aromatics from 7.7 to 33 mass%. Effects of vehicle type and fuel composition on emissions are reported. The results show that emissions of these toxic species were broadly comparable to those observed in previous dynamometer and tunnel studies. Emissions of VOCs and smaller PAHs such as naphthalene, which are derived largely from the combustion process, appear to be related, and show relatively little variability when compared with the variability in emissions of aldehydes and larger PAHs. In particular, aldehyde emissions are highly variable and may be related to engine operating conditions. Fuels of lower sulphur and aromatic content did not have a significant influence on emissions of VOCs and aldehydes, but tended to result in lower emissions of PAHs. The toxicity of vehicle exhaust, as determined by inhalation risk and toxic equivalency factor (TEF)-weighted PAH emissions, was reduced with fuels of lower aromatic content.

  18. Fuel pin bowing and related investigation of the gadolinium fuel pin influence on power release inside of neighbouring fuel pins in a WWER-440 type core

    International Nuclear Information System (INIS)

    Mikus, J.

    2006-01-01

    As known both the WWER-440 and WWER-1000 reactors are systematically modernized to enhance their safety and economical parameters of operation. For this purpose new fuel assemblies (FAs) were designed with improved technical parameters, e.g., containing fuel pins (FPs) in which Gd 2 O 3 burnable absorber is integrated into fuel. Presence of such FPs in reactor core results in a strong depression of thermal neutrons in their positions and corresponding high gradients in neighbouring FPs. Consequently, similar situation in neighbouring FPs can be expected as for both the power release and temperature gradients. The purpose of this work consists in investigation of the gadolinium FP influence on space power distribution, especially from viewpoint of the values and gradient occurrence inside of the neighbouring FPs that could result in static loads with some consequences, e.g., a contribution to FP/FA bowing. Since detailed power distributions cannot be obtained in the NPPs, needed information is provided by means of experiments on research reactors. As for the power release measurement inside of FPs, some special (e.g. track) detectors placed between fuel pellets are usually used. Since such works are relatively complicated and time consuming, an evaluation method based on mathematical modelling and numerical approximation was proposed by means of that, and using measured (integral) power release in selected FPs, needed power release values inside of investigated FPs, can be estimated. For this purpose, experimental results from light water, zero-power research reactor LR-0 obtained by measurements in a WWER-440 type core with 19 FAs at zero boron concentration and containing some FPs with gadolinium (Gd FPs) were utilized. Application of the proposed evaluation method is demonstrated on investigated FPs neighbouring a Gd FP by means of the: relative azimuthal power distribution estimation inside of investigated FPs on their fuel pellet surface in horizontal plane

  19. Preparation of U-Si/U-Me (Me = Fe, Ni, Mn) aluminum-dispersion plate-type fuel (miniplates) for capsule irradiation

    International Nuclear Information System (INIS)

    Ugajin, Mitsuhiro; Itoh, Akinori; Akabori, Mitsuo

    1993-06-01

    Details of equipment installed, method adopted and final products were described on the preparation of uranium silicides and other fuels for capsule irradiation. Main emphasis was placed on the preparation of laboratory-scale aluminum-dispersion plate-type fuel (miniplates) loaded to the first and second JMTR silicide capsules. Fuels contained in the capsules are as follows: (A) uranium-silicide base alloys U 3 Si 2 , Mo- added U 3 Si 2 , U 3 Si 2 +U 3 Si, U 3 Si 2 +USi, U 3 Si, U 3 (Si 0.8 Ge 0.2 ), U 3 (Si 0.6 Ge 0.4 ) (B) U 6 Me-type alloys with higher uranium density U 6 Mn, U 6 Ni, U 6 (Fe 0.4 Ni 0.6 ), U 6 (Fe 0.6 Mn 0.4 ) The powder-metallurgical picture-frame method was adopted and laboratory-scale technique was established for the preparation of miniplates. As a result of inspection for capsule irradiation, miniplates were prepared to meet the requirements of specification. (author)

  20. Metallic Reactor Fuel Fabrication for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Song, Hoon; Kim, Jong-Hwan; Ko, Young-Mo; Woo, Yoon-Myung; Kim, Ki-Hwan; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The metal fuel for an SFR has such advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant, and inherent passive safety 1. U-Zr metal fuel for SFR is now being developed by KAERI as a national R and D program of Korea. The fabrication technology of metal fuel for SFR has been under development in Korea as a national nuclear R and D program since 2007. The fabrication process for SFR fuel is composed of (1) fuel slug casting, (2) loading and fabrication of the fuel rods, and (3) fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycled streams in this fabrication process. Fabrication on the rod type metallic fuel was carried out for the purpose of establishing a practical fabrication method. Rod-type fuel slugs were fabricated by injection casting. Metallic fuel slugs fabricated showed a general appearance was smooth.

  1. Práticas PMBOK® e Corrente Crítica: antagonismos e oportunidades de complementação

    Directory of Open Access Journals (Sweden)

    João Victor Rojas Luiz

    2017-08-01

    Full Text Available Resumo O esforço do Project Management Institute (PMI em reunir e codificar o conhecimento de valor na área de Gerenciamento de Projetos (GP resultou no A Guide to the Project Management Body of Knowledge (Guia PMBOK®. Goldratt contribuiu para o conjunto de conhecimentos em GP ao desenvolver o método Corrente Crítica (Critical Chain Project Management – CCPM, fundamentada na Teoria das Restrições (Theory of Constraints – TOC. A CCPM, superficialmente recomendada pelo Guia PMBOK®, apresenta novos conceitos e métodos voltados ao GP, parte dos quais conflita com algumas das práticas recomendadas pelo próprio Guia. Esta pesquisa assume que a CCPM pode trazer relevantes benefícios ao campo de conhecimento em GP e questiona a pouca relevância dada pelo Guia PMBOK®. Assim, a pesquisa tem como proposta global verificar inicialmente o grau de inserção da CCPM no Guia e, posteriormente, avaliar possíveis antagonismos entre eles, assim como oportunidades de complementação. A análise se baseou em uma revisão da literatura e em uma pesquisa de campo com especialistas certificados. Apesar de a CCPM ser recomendada pelo Guia como um método voltado para o Gerenciamento do Tempo, a pesquisa indicou que outras práticas que envolvem a CCPM, e não contempladas pelo Guia, podem contribuir não apenas para o Gerenciamento do Tempo, mas também de Recursos Humanos e Comunicações. A pesquisa apontou também que a CCPM não é autossuficiente em suas práticas, devendo recorrer ao Guia especialmente no Gerenciamento da Integração e Escopo. Foi indicado ainda que a CCPM se opõe a certas práticas de gestão do tempo recomendadas pelo Guia. Se aplicadas concomitantemente, elas podem potencialmente pôr em risco a efetividade do GP.

  2. Effects of Cooling Fluid Flow Rate on the Critical Heat Flux and Flow Stability in the Plate Fuel Type 2 MW TRIGA Reactor

    OpenAIRE

    H. P. Rahardjo; V. I. Sri Wardhani

    2017-01-01

    The conversion program of the 2 MW TRIGA reactor in Bandung consisted of the replacement of cylindrical fuel (produced by General Atomic) with plate fuel (produced by BATAN). The replacement led into the change of core cooling process from upward natural convection type to downward forced convection type, and resulted in different thermohydraulic safety criteria, such as critical heat flux (CHF) limit, boiling limit, and cooling fluid flow stability. In this paper, a thermohydraulic safety an...

  3. Advanced Research Reactor Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C. K.; Park, H. D.; Kim, K. H. (and others)

    2006-04-15

    RERTR program for non-proliferation has propelled to develop high-density U-Mo dispersion fuels, reprocessable and available as nuclear fuel for high performance research reactors in the world. As the centrifugal atomization technology, invented in KAERI, is optimum to fabricate high-density U-Mo fuel powders, it has a great possibility to be applied in commercialization if the atomized fuel shows an acceptable in-reactor performance in irradiation test for qualification. In addition, if rod-type U-Mo dispersion fuel is developed for qualification, it is a great possibility to export the HANARO technology and the U-Mo dispersion fuel to the research reactors supplied in foreign countries in future. In this project, reprocessable rod-type U-Mo test fuel was fabricated, and irradiated in HANARO. New U-Mo fuel to suppress the interaction between U-Mo and Al matrix was designed and evaluated for in-reactor irradiation test. The fabrication process of new U-Mo fuel developed, and the irradiation test fuel was fabricated. In-reactor irradiation data for practical use of U-Mo fuel was collected and evaluated. Application plan of atomized U-Mo powder to the commercialization of U-Mo fuel was investigated.

  4. Development of the uranium recovery process from rejected fuel plates in the fabrication of MTR type nuclear fuel

    International Nuclear Information System (INIS)

    Fleming Rubio, Peter Alex

    2010-01-01

    The current work was made in Conversion laboratory belonging to Chilean Nuclear Energy Commission, CCHEN. This is constituted by the development of three hydrometallurgical processes, belonging to the recovery of uranium from fuel plates based on uranium silicide (U_3Si_2) process, for nuclear research reactors MTR (Material Testing Reactor) type, those that come from the Fuel Elements Manufacture Plant, PEC. In the manufacturing process some of these plates are subjected to destructive tests by quality requirement or others are rejected for non-compliance with technical specifications, such as: lack of homogenization of the dispersion of uraniferous compound in the meat, as well as the appearance of the defects, such as blisters, so-called "dog bone", "fish tail", "remote islands", among others. Because the uranium used is enriched in 19.75% U_2_3_5 isotope, which explains the high value in the market, it must be recovered for reuse, returning to the production line of fuel elements. The uranium silicide, contained in the plates, is dispersed in an aluminum matrix and covered with plates and frames of ASTM 6061 Aluminum, as a sandwich coating, commonly referred to as 'meat' (sandwich meat). As aluminum is the main impurity, the process begins with this metal dissolution, present in meat and plates, by NaOH reaction, followed by a vacuum filtration, washing and drying, obtaining a powder of uranium silicide, with a small impurities percentage. Then, the crude uranium silicide reacts with a solution of hydrofluoric acid, dissolving the silicon and simultaneously precipitating UF_4 by reaction with HNO_3, obtaining an impure UO_2(NO_3)_2 solution. The experimental work was developed and implemented at laboratory scale for the three stages pertaining to the uranium recovery process, determining for each one the optimum operation conditions: temperature, molarity or concentration, reagent excess, among others (author)

  5. Nuclear fuel

    International Nuclear Information System (INIS)

    Azevedo, J.B.L. de.

    1980-01-01

    All stages of nuclear fuel cycle are analysed with respect to the present situation and future perspectives of supply and demand of services; the prices and the unitary cost estimation of these stages for the international fuel market are also mentioned. From the world resources and projections of uranium consumption, medium-and long term analyses are made of fuel availability for several strategies of use of different reactor types. Finally, the cost of nuclear fuel in the generation of electric energy is calculated to be used in the energetic planning of the electric sector. (M.A.) [pt

  6. Outward electron transfer by Saccharomyces cerevisiae monitored with a bi-cathodic microbial fuel cell-type activity sensor.

    Science.gov (United States)

    Ducommun, Raphaël; Favre, Marie-France; Carrard, Delphine; Fischer, Fabian

    2010-03-01

    A Janus head-like bi-cathodic microbial fuel cell was constructed to monitor the electron transfer from Saccharomyces cerevisiae to a woven carbon anode. The experiments were conducted during an ethanol cultivation of 170 g/l glucose in the presence and absence of yeast-peptone medium. First, using a basic fuel-cell type activity sensor, it was shown that yeast-peptone medium contains electroactive compounds. For this purpose, 1% solutions of soy peptone and yeast extract were subjected to oxidative conditions, using a microbial fuel cell set-up corresponding to a typical galvanic cell, consisting of culture medium in the anodic half-cell and 0.5 M K(3)Fe(CN)(6) in the cathodic half-cell. Second, using a bi-cathodic microbial fuel cell, it was shown that electrons were transferred from yeast cells to the carbon anode. The participation of electroactive compounds in the electron transport was separated as background current. This result was verified by applying medium-free conditions, where only glucose was fed, confirming that electrons are transferred from yeast cells to the woven carbon anode. Knowledge about the electron transfer through the cell membrane is of importance in amperometric online monitoring of yeast fermentations and for electricity production with microbial fuel cells. Copyright (c) 2009 John Wiley & Sons, Ltd.

  7. Neutronic, thermal-hydraulics and safety calculations of a Miniplate Irradiation Device (MID) of dispersion type fuel elements

    International Nuclear Information System (INIS)

    Domingos, Douglas Borges

    2010-01-01

    Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of a Miniplate Irradiation Device (MID) to be placed in the IEA-R1 reactor core. The irradiation device is used to receive miniplates of U 3 O 8 -Al and U 3 Si 2 - Al dispersion fuels, LEU type (19.75 % 235 U) with uranium densities of, respectively, 3.2 gU/cm 3 and 4.8 gU/cm 3 . The fuel miniplates will be irradiated to nominal 235 U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor (RMB), now in the conception phase. For the neutronic calculation, the computer codes CITATION and 2DB were utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer codes LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. The calculations showed that the irradiation should occur without adverse consequences in the IEA-R1 reactor. (author)

  8. Fuel assembly

    International Nuclear Information System (INIS)

    Bando, Masaru.

    1993-01-01

    As neutron irradiation progresses on a fuel assembly of an FBR type reactor, a strong force is exerted to cause ruptures if the arrangement of fuel elements is not displaced, whereas the fuel elements may be brought into direct contact with each other not by way of spacers to cause burning damages if the arrangement is displaced. In the present invention, the circumference of fuel elements arranged in a normal triangle lattice is surrounded by a wrapper tube having a hexagonal cross section, wire spacers are wound therearound, and deformable spacers are distributed to optional positions for fuel elements in the wrapper tube. Interaction between the fuel elements caused by irradiation is effectively absorbed, thereby enabling to delay the occurrence of the rupture and burning damages of the elements. (N.H.)

  9. Fuel Cells

    DEFF Research Database (Denmark)

    Smith, Anders; Pedersen, Allan Schrøder

    2014-01-01

    Fuel cells have been the subject of intense research and development efforts for the past decades. Even so, the technology has not had its commercial breakthrough yet. This entry gives an overview of the technological challenges and status of fuel cells and discusses the most promising applications...... of the different types of fuel cells. Finally, their role in a future energy supply with a large share of fluctuating sustainable power sources, e.g., solar or wind, is surveyed....

  10. Multi-fuel reformers for fuel cells used in transportation. Phase 1: Multi-fuel reformers

    Science.gov (United States)

    1994-05-01

    DOE has established the goal, through the Fuel Cells in Transportation Program, of fostering the rapid development and commercialization of fuel cells as economic competitors for the internal combustion engine. Central to this goal is a safe feasible means of supplying hydrogen of the required purity to the vehicular fuel cell system. Two basic strategies are being considered: (1) on-board fuel processing whereby alternative fuels such as methanol, ethanol or natural gas stored on the vehicle undergo reformation and subsequent processing to produce hydrogen, and (2) on-board storage of pure hydrogen provided by stationary fuel processing plants. This report analyzes fuel processor technologies, types of fuel and fuel cell options for on-board reformation. As the Phase 1 of a multi-phased program to develop a prototype multi-fuel reformer system for a fuel cell powered vehicle, the objective of this program was to evaluate the feasibility of a multi-fuel reformer concept and to select a reforming technology for further development in the Phase 2 program, with the ultimate goal of integration with a DOE-designated fuel cell and vehicle configuration. The basic reformer processes examined in this study included catalytic steam reforming (SR), non-catalytic partial oxidation (POX) and catalytic partial oxidation (also known as Autothermal Reforming, or ATR). Fuels under consideration in this study included methanol, ethanol, and natural gas. A systematic evaluation of reforming technologies, fuels, and transportation fuel cell applications was conducted for the purpose of selecting a suitable multi-fuel processor for further development and demonstration in a transportation application.

  11. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Makoto; Ogiya, Shunsuke.

    1989-01-01

    For improving the economy of a BWR type reactor by making the operation cycle longer, the fuel enrichment degree has to be increased further. However, this makes the subcriticality shallower in the upper portion of the reactor core, to bring about a possibility that the reactor shutdown becomes impossible. In the present invention, a portion of fuel rod is constituted as partial length fuel rods (P-fuel rods) in which the entire stack length in the effective portion is made shorter by reducing the concentration of fissionable materials in the axial portion. A plurality of moderator rods are disposed at least on one diagonal line of a fuel assembly and P-fuel rods are arranged at a position put between the moderator rods. This makes it possible to reactor shutdown and makes the axial power distribution satisfactory even if the fuel enrichment degree is increased. (T.M.)

  12. Fuel Services

    International Nuclear Information System (INIS)

    Silberstein, A.

    1982-09-01

    FRAGEMA has developed most types of inspection equipments to work on irradiated fuel assemblies and on single fuel rods during reactor outages with an efficiency compatible with the utilities operating priorities. In order to illustrate this statement, two specific examples of inspection equipments are shortly described: the on-site removable fuel rod assembly examination stand, and the fuel assembly multiple examination device. FRAGEMA has developed techniques for the identifiction of the leaking fuel rods in the fuel assembly and the tooling necessary to perform the replacement of the faulted element. These examples of methods, techniques and equipments described and the experience accumulated through their use allow FRAGEMA to qualify for offering the supply of the corresponding software, hardware or both whenever an accurate understanding of the fuel behaviour is necessary and whenever direct intervention on the assembly and associated components is necessary due to safety, operating or economical reasons

  13. Vibration characteristics of a PWR fuel rod supported by optimized H type spacer grids

    International Nuclear Information System (INIS)

    Choi, M. H.; Kang, H. S.; Yoon, K. H.; Kim, H. K.; Song, K. N.

    2002-01-01

    The spacer grids are one of the main structural components in the fuel assembly, which supports and protects the fuel rods from the external loads by seismic and coolant flow. In this study, a modal test and a FE vibration analysis using ABAQUS are performed on a PWR dummy fuel rod of 3.847 m which is continuously supported by eight Optimized H type spacer grids. The experimental results agree with previous works that the natural frequencies decrease, while the amplitudes increase, with the increase of the excitation force. The force levels showing the maximum displacement of 0.2 mm are in the range from 0.2 N to 0.3 N, and at the same force range the fundamental frequencies are measured around 42.0 Hz, at which the relatively big displacements are observed at the 7th span. The results from the modal tests and the FE analyses are compared by both Modal Assurance Criteria (MAC) values and mode shapes. The MAC values at 2nd, 4th, and 7th mode are below 50%. It is believed that the reason of the low MACs at those modes is that the vibration amplitudes of the modes are more distorted by the excitation force than those of the other modes

  14. Fuels characterization studies. [jet fuels

    Science.gov (United States)

    Seng, G. T.; Antoine, A. C.; Flores, F. J.

    1980-01-01

    Current analytical techniques used in the characterization of broadened properties fuels are briefly described. Included are liquid chromatography, gas chromatography, and nuclear magnetic resonance spectroscopy. High performance liquid chromatographic ground-type methods development is being approached from several directions, including aromatic fraction standards development and the elimination of standards through removal or partial removal of the alkene and aromatic fractions or through the use of whole fuel refractive index values. More sensitive methods for alkene determinations using an ultraviolet-visible detector are also being pursued. Some of the more successful gas chromatographic physical property determinations for petroleum derived fuels are the distillation curve (simulated distillation), heat of combustion, hydrogen content, API gravity, viscosity, flash point, and (to a lesser extent) freezing point.

  15. In-pile tests of HTGR fuel particles and fuel elements

    International Nuclear Information System (INIS)

    Chernikov, A.S.; Kolesov, V.S.; Deryugin, A.I.

    1985-01-01

    Main types of in-pile tests for specimen tightness control at the initial step, research of fuel particle radiation stability and also study of fission product release from fuel elements during irradiation are described in this paper. Schemes and main characteristics of devices used for these tests are also given. Principal results of fission gas product release measurements satisfying HTGR demands are illustrated on the example of fuel elements, manufactured by powder metallurgy methods and having TRISO fuel particles on high temperature pyrocarbon and silicon carbide base. (author)

  16. Advanced fuel in the Budapest research reactor

    International Nuclear Information System (INIS)

    Hargitai, T.; Vidovsky, I.

    1997-01-01

    The Budapest Research Reactor, the first nuclear facility of Hungary, started to operate in 1959. The main goal of the reactor is to serve neutron research, but applications as neutron radiography, radioisotope production, pressure vessel surveillance test, etc. are important as well. The Budapest Research Reactor is a tank type reactor, moderated and cooled by light water. After a reconstruction and upgrading in 1967 the VVR-SM type fuel elements were used in it. These fuel elements provided a thermal power of 5 MW in the period 1967-1986 and 10 MW after the reconstruction from 1992. In the late eighties the Russian vendor changed the fuel elements slightly, i.e. the main parameters of the fuel remained unchanged, however a higher uranium content was reached. This new fuel is called VVR-M2. The geometry of VVR-SM and VVR-M2 are identical, allowing the use to load old and new fuel assemblies together to the active core. The first new type fuel assemblies were loaded to the Budapest Research Reactor in 1996. The present paper describes the operational experience with the new type of fuel elements in Hungary. (author)

  17. Nuclear fuel

    International Nuclear Information System (INIS)

    Quinauk, J.P.

    1990-01-01

    Since 1985, Fragema has been marketing and selling the Advanced Fuel Assemby AFA whose main features are its zircaloy grids and removable top and bottom nozzles. It is this product, which exists for several different fuel assembly arrays and heights, that will be employed in the reactors at Daya Bay. Fragema employs gadolinium as the consumable poison to enable highperformance fuel management. More recently, the company has supplied fuel assemblies of the mixed-oxide(MOX) and enriched reprocessed uranium type. The reliability level of the fuel sold by Fragema is one of the highest in the world, thanks in particular to the excellence of the quality assurance and quality control programs that have been implemented at all stages of its design and manufacture

  18. The analysis of effect of gaseous fission products on service-ability of various types of fuel elements used in research reactors

    International Nuclear Information System (INIS)

    Vatulin, A.A.; Kulakov, G.V.; Kosaurov, A.A.; Dobrikova, I.V.; Morozov, A.A.

    2005-01-01

    The analysis of the work to license U-Mo dispersion fuel that has been carried out by program RERTR shows that at high burnups the U-Mo particles interacts with an Al matrix to form porosity in the central areas of the fuel meat. This results leads to change the loading of cladding due to the formation of long-length defects in the fuel meat. As a result, the pressure of gases leads to a larger shape change (pillowing) of fuels and in some instances to their failure. The paper presents the results of modeling pillowing process proceeding in Russia's research reactor fuels of different types (tubular and rod fuels) that is effected by the accumulation and pressure of fission gas in meat porosity. The problem has been resolved in the viscous-plasticity statement using the program MARC for finite element analysis. The behavior of fission gas (FG) in open porosity has been analyzed to reveal the influence of the fuel design on their resistance to the pressure of FG. (author)

  19. LEU fuel fabrication in Argentina

    International Nuclear Information System (INIS)

    Giorsetti, D.R.; Gomez, J.O.; Marajofsky, A.; Kohut, C.

    1985-01-01

    As an Institution, aiming to meet with its own needs, CNEA has been intensively developing reduced enriched fuel to use in its own research and test reactors. Development of the fabrication technology as well as the design, installation and operation of the manufacturing plant, have been carried out with its own funds. Irradiation and post-irradiation of test miniplates have been taking place within the framework of the RERTR program. During the last years, CNEA has developed three LEU fuel types. In the previous RERTR meetings, we presented the technological results obtained with these fuel types. This paper focuses on CNEA LEU fuel element manufacturing status and the trained personnel we can offer in design and manufacture fuel capability. CNEA has its own fuel manufacturing technology; the necessary facilities to start the fuel fabrication; qualified technicians and professionals for: fuel design and behaviour analysis; fuel manufacturing and QA; international recognition of its fuel development and manufacturing capability through its ORR miniplate irradiation; its own natural uranium and the future possibility to enrich up to 20% U 235 ; the probability to offer a competitive fuel manufacturing cost in the international market; the disposition to cooperate with all countries that wish to take part and aim to reach an self-sufficiency in their own fuel supply needs

  20. BWR type nuclear reactors

    International Nuclear Information System (INIS)

    Yamamoto, Toru.

    1987-01-01

    Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

  1. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, New York (United States)

    2007-07-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat.

  2. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan; Kim, Yeon Soo

    2007-01-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

  3. Modelling the effects of transport policy levers on fuel efficiency and national fuel consumption

    International Nuclear Information System (INIS)

    Kirby, H.R.; Hutton, B.; McQuaid, R.W.; Napier Univ., Edinburgh; Raeside, R.; Napier Univ., Edinburgh; Zhang, Xiayoan; Napier Univ., Edinburgh

    2000-01-01

    The paper provides an overview of the main features of a Vehicle Market Model (VMM) which estimates changes to vehicle stock/kilometrage, fuel consumed and CO 2 emitted. It is disaggregated into four basic vehicle types. The model includes: the trends in fuel consumption of new cars, including the role of fuel price: a sub-model to estimate the fuel consumption of vehicles on roads characterised by user-defined driving cycle regimes; procedures that reflect distribution of traffic across different area/road types; and the ability to vary the speed (or driving cycle) from one year to another, or as a result of traffic growth. The most significant variable influencing fuel consumption of vehicles was consumption in the previous year, followed by dummy variables related to engine size. the time trend (a proxy for technological improvements), and then fuel price. Indeed the effect of fuel price on car fuel efficiency was observed to be insignificant (at the 95% level) in two of the three versions of the model, and the size of fuel price term was also the smallest. This suggests that the effectiveness of using fuel prices as a direct policy tool to reduce fuel consumption may he limited. Fuel prices may have significant indirect impacts (such as influencing people to purchase more fuel efficient cars and vehicle manufacturers to invest in developing fuel efficient technology) as may other factors such as the threat of legislation. (Author)

  4. Evolution of Particle Bed Reactor Fuel

    Science.gov (United States)

    Jensen, Russell R.; Evans, Robert S.; Husser, Dewayne L.; Kerr, John M.

    1994-07-01

    To realize the potential performance advantages inherent in a particle bed reactor (PBR) for nuclear thermal propulsion (NTP) applications, high performance particle fuel is required. This fuel must operate safely and without failure at high temperature in high pressure, flowing hydrogen propellant. The mixed mean outlet temperature of the propellant is an important characteristic of PBR performance. This temperature is also a critical parameter for fuel particle design because it dictates the required maximum fuel operating temperature. In this paper, the evolution in PBR fuel form to achieve higher operating temperatures is discussed and the potential thermal performance of the different fuel types is evaluated. It is shown that the optimum fuel type for operation under the demanding conditions in a PBR is a coated, solid carbide particle.

  5. Detection of delamination defects in plate type fuel elements applying an automated C-Scan ultrasonic system

    International Nuclear Information System (INIS)

    Katchadjian, P.; Desimone, C.; Ziobrowski, C.; Garcia, A.

    2002-01-01

    For the inspection of plate type fuel elements to be used in Research Nuclear Reactors it was applied an immersion pulse-echo ultrasonic technique. For that reason an automated movement system was implemented according to the axes X, Y and Z that allows to automate the test and to show the results obtained in format of C-Scan, facilitating the immediate identification of possible defects and making repetitive the inspection. In this work problems found during the laboratory tests and factors that difficult the inspection are commented. Also the results of C-Scans over UMo fuel elements with pattern defects are shown. Finally, the main characteristics of the transducer with the one the better results were obtained are detailed. (author)

  6. Simplified CFD model of coolant channels typical of a plate-type fuel element: an exhaustive verification of the simulations

    Energy Technology Data Exchange (ETDEWEB)

    Mantecón, Javier González; Mattar Neto, Miguel, E-mail: javier.mantecon@ipen.br, E-mail: mmattar@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    The use of parallel plate-type fuel assemblies is common in nuclear research reactors. One of the main problems of this fuel element configuration is the hydraulic instability of the plates caused by the high flow velocities. The current work is focused on the hydrodynamic characterization of coolant channels typical of a flat-plate fuel element, using a numerical model developed with the commercial code ANSYS CFX. Numerical results are compared to accurate analytical solutions, considering two turbulence models and three different fluid meshes. For this study, the results demonstrated that the most suitable turbulence model is the k-ε model. The discretization error is estimated using the Grid Convergence Index method. Despite its simplicity, this model generates precise flow predictions. (author)

  7. Simplified CFD model of coolant channels typical of a plate-type fuel element: an exhaustive verification of the simulations

    International Nuclear Information System (INIS)

    Mantecón, Javier González; Mattar Neto, Miguel

    2017-01-01

    The use of parallel plate-type fuel assemblies is common in nuclear research reactors. One of the main problems of this fuel element configuration is the hydraulic instability of the plates caused by the high flow velocities. The current work is focused on the hydrodynamic characterization of coolant channels typical of a flat-plate fuel element, using a numerical model developed with the commercial code ANSYS CFX. Numerical results are compared to accurate analytical solutions, considering two turbulence models and three different fluid meshes. For this study, the results demonstrated that the most suitable turbulence model is the k-ε model. The discretization error is estimated using the Grid Convergence Index method. Despite its simplicity, this model generates precise flow predictions. (author)

  8. SR 97: Post-closure safety for a KBS 3 type deep repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    Hedin, A.; Kautsky, U.

    2000-03-01

    Prior to coming site investigations for siting of a deep repository for spent nuclear fuel, SKB has carried out the long-term safety assessment SR 97, requested by the Swedish Government. The repository is of the KBS-3 type, where the fuel is placed in isolating copper canisters with a high-strength cast iron insert. The canisters are surrounded by bentonite clay in individual deposition holes at a depth of 500 m in granitic bedrock. The future evolution of the repository system is analysed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings, including climate, persist. The four other scenarios show the evolution if the repository contains a few initially defective canisters, in the event of climate change, in the event of earthquakes, and in the event of future inadvertent human intrusion. The principal conclusion of the assessment is that the prospects of building a safe deep repository for spent nuclear fuel in Swedish granitic bedrock are very good. (author)

  9. Nuclear power and its fuel cycle

    International Nuclear Information System (INIS)

    Wymer, R.G.

    1986-01-01

    A series of viewgraphs describes the nuclear fuel cycle and nuclear power, covering reactor types, sources of uranium, enrichment of uranium, fuel fabrication, transportation, fuel reprocessing, and radioactive wastes

  10. Selection of the new position for the instrumented fuel Flip type; Seleccion de la nueva posicion para el combustible instrumentado tipo Flip

    Energy Technology Data Exchange (ETDEWEB)

    Paredes G, L C

    1991-12-15

    It was selected and proved the new position for the instrumented fuel, in function of the smallest influence on the temperature of the fuel due to movements of the control bars, as well as to its functionality regarding their geometric localization in the core. Therefore, it intends that the instrumented fuel Flip type, pass from the position C-9 to the C-6. It was selected the turn angle of the fuel of 180 C, for to be the one that presents a smaller influence of the control bars and it allows to register the values of temperature but high. It was as the movements of the control bars they produce local interferences of the neutron flow in the fuels and that in the case of the ring C, these interferences became maximum variations of the fuel temperature of 125 C, for operations to 1 MW in the thermal column. (Author)

  11. Sex-related differences in fuel utilization and hormonal response to exercise: implications for individuals with type 1 diabetes.

    Science.gov (United States)

    Brockman, Nicole K; Yardley, Jane E

    2018-06-01

    Sex-related differences in metabolic and neuroendocrine response to exercise in individuals without diabetes have been well established. Men and women differ in fuel selection during exercise, in which women rely to a greater extent on fat oxidation, whereas males rely mostly on carbohydrate oxidation for energy production. The difference in fuel selection appears to be mediated by sex-related differences in hormonal (including catecholamines, growth hormone, and estrogen) response to different types and intensities of exercise. In general, men exhibit an amplified counter-regulatory response to exercise, with elevated levels of catecholamines compared with women. However, women exhibit greater sensitivity to the lipolytic action of the catecholamines and deplete less of their glycogen stores than men during exercise, which suggests that women may experience a greater defense in blood glucose control after exercise than men. Conversely, little is known about sex-related differences in response to exercise in individuals with type 1 diabetes (T1D). A single study investigating sex-related differences in response to moderate aerobic exercise in individuals with T1D found sex-related differences in catecholamine response and fuel selection, but changes in blood glucose were not measured. To our knowledge, there are no studies investigating sex-related differences in blood glucose responses to different types and intensities of exercise in individuals with T1D. This review summarizes sex-related differences in exercise responses that could potentially impact blood glucose levels during exercise in individuals with T1D and highlights the need for further research.

  12. Reducing Fuel Volatility. An Additional Benefit From Blending Bio-fuels?

    Energy Technology Data Exchange (ETDEWEB)

    Bailis, R. [Yale School of Forestry and Environmental Studies, 195 Prospect Street, New Haven, CT 06511 (United States); Koebl, B.S. [Utrecht University, Science Technology and Society, Budapestlaan 6, 3584 CD Utrecht (Netherlands); Sanders, M. [Utrecht University, Utrecht School of Economics, Janskerkhof 12, 3512 BL Utrecht (Netherlands)

    2011-02-15

    Oil price volatility harms economic growth. Diversifying into different fuel types can mitigate this effect by reducing volatility in fuel prices. Producing bio-fuels may thus have additional benefits in terms of avoided damage to macro-economic growth. In this study we investigate trends and patterns in the determinants of a volatility gain in order to provide an estimate of the tendency and the size of the volatility gain in the future. The accumulated avoided loss from blending gasoline with 20 percent ethanol-fuel estimated for the US economy amounts to 795 bn. USD between 2010 and 2019 with growing tendency. An amount that should be considered in cost-benefit analysis of bio-fuels.

  13. Study on the inspection item and inspection method of HTGR fuel

    International Nuclear Information System (INIS)

    Na, Sang Ho; Kim, Y. K.; Jeong, K. C.; Oh, S. C.; Cho, M. S.; Kim, Y. M.; Lee, Y. W.

    2006-01-01

    The type of HTGR(High Temperature Gas-cooled Reactor) fuel is different according to the reactor type. Generally the HTGR fuel has two types. One is a block type, which is manufactured in Japan or America. And the other is a pebble type, which is manufactured in China. Regardless of the fuel type, the fuel manufacturing process started from the coated particle, which is consisted of fuel kernel and the 4 coating layers. Korea has a plan to fabricate a HTGR fuel in near future. The appropriate quality inspection standards are requested to produce a sound and reliable coated particle for HTGR fuel. Therefore, the inspection items and the inspection methods of HTGR fuel between Japan and China, which countries have the manufacturing process, are investigated to establish a proper inspection standards of our product characteristics

  14. Inter renewal travelling wave reactor with rotary fuel columns

    International Nuclear Information System (INIS)

    Terai, Yuzo

    2016-01-01

    To realize the COP21 decision, this paper proposes Inter Renewal Travelling Wave Reactor that bear high burn-up rate 50% and product TRU fuel efficiently. The reactor is based on 4S Fast Reactor and has Reactor Fuel Columns as fuel assemblies that equalize temperature in the fuel assembly so that fewer structure is need to restrain thermal transformation. To equalize burn-up rate of all fuel assemblies in the reactor, each rotary fuel column has each motor-lifter. The rotary fuel column has two types (Cylinder type and Heat Pipe type using natrium at 15 kPa which supply high temperature energy for Ultra Super Critical power plant). At 4 years cycle all rotary fuel columns of the reactor are renewed by the metallurgy method (vacuum re-smelting) and TRU fuel is gotten from the water fuel. (author)

  15. Fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Aoyama, Motoo; Koyama, Jun-ichi; Uchikawa, Sadao; Bessho, Yasunori; Nakajima, Akiyoshi; Maruyama, Hiromi; Ozawa, Michihiro; Nakamura, Mitsuya.

    1990-01-01

    The present invention concerns fuel assemblies charged in a BWR type reactor and the reactor core. The fuel assembly comprises fuel rods containing burnable poisons and fuel rods not containing burnable poisons. Both of the highest and the lowest gadolinia concentrations of the fuel rods containing gadolinia as burnable poisons are present in the lower region of the fuel assembly. This can increase the spectral shift effect without increasing the maximum linear power density. (I.N.)

  16. Multi-objective regulations on transportation fuels: Comparing renewable fuel mandates and emission standards

    International Nuclear Information System (INIS)

    Rajagopal, D.; Plevin, R.; Hochman, G.; Zilberman, D.

    2015-01-01

    We compare two types of fuel market regulations — a renewable fuel mandate and a fuel emission standard — that could be employed to simultaneously achieve multiple outcomes such as reduction in fuel prices, fuel imports and greenhouse gas (GHG) emissions. We compare these two types of regulations in a global context taking into account heterogeneity in carbon content of both fossil fuels and renewable fuels. We find that although neither the ethanol mandate nor the emission standard is certain to reduce emissions relative to a business-as-usual baseline, at any given level of biofuel consumption in the policy region, a mandate, relative to an emission standard, results in higher GHG emissions, smaller expenditure on fuel imports, lower price of ethanol-blended gasoline and higher domestic fuel market surplus. This result holds over a wide range of values of model parameters. We also discuss the implications of this result to a regulation such as the US Renewable Fuel Standard given recent developments within the US such as increase in shale and tight oil production and large increase in average vehicle fuel economy of the automotive fleet. - Highlights: • Biofuel mandates and fuel GHG emission standards are analyzed from a multiple criteria perspective • An emission-standard always results in lower global emissions while requiring less biofuel relative to a biofuel mandate • An emission-standard results in higher fuel price in the home region relative to a biofuel mandate • Emission standards lead to more shuffling of both fossil fuels and biofuels between home and abroad • The relative impact of the policies on fuel imports depends on the relative cost-effectiveness of domestic & imported biofuel • Recent developments oil production and fuel economy increase the net benefits of an LCFS approach relative to RFS

  17. Water reactor fuel activities in Russia

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, N [State Scientific Centre of Russian Federation, A.A Bochvar All-Russian Research Inst. of Inorganic Materials, Moscow (Russian Federation)

    1997-12-01

    The presentation reviews the following issues: some specific features of Russian WWER type fuel assemblies and fuel rods; WWER fuel performance; fuel status after irradiation; main directions of programme towards high burnup; development of absorber element. 8 refs, 13 figs, 3 tabs.

  18. Water reactor fuel activities in Russia

    International Nuclear Information System (INIS)

    Sokolov, N.

    1997-01-01

    The presentation reviews the following issues: some specific features of Russian WWER type fuel assemblies and fuel rods; WWER fuel performance; fuel status after irradiation; main directions of programme towards high burnup; development of absorber element. 8 refs, 13 figs, 3 tabs

  19. 3-Dimensional Computational Fluid Dynamics Modeling of Solid Oxide Fuel Cell Using Different Fuels

    Science.gov (United States)

    2011-01-01

    major types of fuel cells in practice are listed below: Polymer Electrolyte Membrane Fuel Cell ( PEMFC ) Alkaline Fuel cell (AFC) Phosphoric Acid...Material Operating Temperature (oC) Efficiency (%) PEMFC H2, Methanol, Formic Acid Hydrated Organic Polymer < 90 40-50 AFC Pure H2 Aqueous

  20. Development of spent fuel dry storage technology

    International Nuclear Information System (INIS)

    Maruoka, Kunio; Matsunaga, Kenichi; Kunishima, Shigeru

    2000-01-01

    The spent fuels are the recycle fuel resources, and it is very important to store the spent fuels in safety. There are two types of the spent fuel interim storage system. One is wet storage system and another is dry storage system. In this study, the dry storage technology, dual purpose metal cask storage and canister storage, has been developed. For the dual purpose metal cask storage, boronated aluminum basket cell, rational cask body shape and shaping process have been developed, and new type dual purpose metal cask has been designed. For the canister storage, new type concrete cask and high density vault storage technology have been developed. The results of this study will be useful for the spent fuel interim storage. Safety and economical spent fuel interim storage will be realized in the near future. (author)

  1. Experience with nuclear fuel utilization in Bulgaria

    Energy Technology Data Exchange (ETDEWEB)

    Harizanov, Y [Committee on the Use of Atomic Energy for Peaceful Purposes, Sofia (Bulgaria)

    1997-12-01

    The presentation on experience with nuclear fuel utilization in Bulgaria briefly reviews the situation with nuclear energy in Bulgaria and then discusses nuclear fuel performance (amount of fuel loaded, type of fuel, burnup, fuel failures, assemblies deformation). 2 tabs.

  2. Fuel cells - a perspective

    International Nuclear Information System (INIS)

    Biegler, T.

    2005-01-01

    Unfortunately, fuel cell publicity conveys expectations and hopes that are often based on uncritical interpretations of the underlying science. The aim here is to use that science to analyse how the technology has developed and what can realistically be delivered by fuel cells. There have been great achievements in fuel cell technology over the past decade, with most types reaching an advanced stage of engineering development. But there has been some muddled thinking about one critical aspect, fuel cell energy efficiency. The 'Carnot cycle' argument, that fuel cells must be much more efficient than heat engines, is a red herring, of no help in predicting real efficiencies. In practice, fuel cells are not always particularly efficient and there are good scientific reasons for this. Cost reduction is a big issue for fuel cells. They are not in principle especially simple devices. Better engineering and mass production will presumably bring costs down, but because of their inherent complexity there is no reason to expect them to be cheap. It is fair to conclude that predictions of fuel cells as commonplace components of energy systems (including a hydrogen economy) need to be treated with caution, at least until major improvements eventuate. However, one type, the direct methanol fuel cell, is aimed at a clear existing market in consumer electronics

  3. Limitations of Commercializing Fuel Cell Technologies

    Science.gov (United States)

    Nordin, Normayati

    2010-06-01

    Fuel cell is the technology that, nowadays, is deemed having a great potential to be used in supplying energy. Basically, fuel cells can be categorized particularly by the kind of employed electrolyte. Several fuel cells types which are currently identified having huge potential to be utilized, namely, Solid Oxide Fuel Cells (SOFC), Molten Carbonate Fuel Cells (MCFC), Alkaline Fuel Cells (AFC), Phosphoric Acid Fuel Cells (PAFC), Polymer Electron Membrane Fuel Cell (PEMFC), Direct Methanol Fuel Cells (DMFC) and Regenerative Fuel Cells (RFC). In general, each of these fuel cells types has their own characteristics and specifications which assign the capability and suitability of them to be utilized for any particular applications. Stationary power generations and transport applications are the two most significant applications currently aimed for the fuel cell market. It is generally accepted that there are lots of advantages if fuel cells can be excessively commercialized primarily in context of environmental concerns and energy security. Nevertheless, this is a demanding task to be accomplished, as there is some gap in fuel cells technology itself which needs a major enhancement. It can be concluded, from the previous study, cost, durability and performance are identified as the main limitations to be firstly overcome in enabling fuel cells technology become viable for the market.

  4. Fuel charging machine

    International Nuclear Information System (INIS)

    Uchikawa, Sadao.

    1978-01-01

    Purpose: To enable continuous fuel discharging and charging steps in a bwr type reactor by effecting positioning only for once by providing a plurality of fuel assembly grippers and their drives co-axially on a rotatable surface. Constitution: A plurality of fuel assembly grippers and their drives are provided co-axially on a rotatable surface. For example, a gripper A, a drive B, a gripper C and a drive D are arranged co-axially in symmetric positions on a disk rotated on rails by wheels and rotational drives. A new fuel in a fuel pool is gripped by the gripper A and transported above the reactor core. Then, the disk is positioned so that the gripper C can grip the spent fuel in the core, and the fuel to be discharged is gripped and raised by the gripper C. Then the disk is rotated by 180 0 and the new fuel in the gripper A is charged into the position from which the old fuel has been discharged and, finally, the discharged fuel is sent to the fuel pool for storage. (Seki, T.)

  5. Comprehensive Analysis of Trends and Emerging Technologies in All Types of Fuel Cells Based on a Computational Method

    Directory of Open Access Journals (Sweden)

    Takaya Ogawa

    2018-02-01

    Full Text Available Fuel cells have been attracting significant attention recently as highly efficient and eco-friendly energy generators. Here, we have comprehensively reviewed all types of fuel cells using computational analysis based on a citation network that detects emerging technologies objectively and provides interdisciplinary data to compare trends. This comparison shows that the technologies of solid oxide fuel cells (SOFCs and electrolytes in polymer electrolyte fuel cells (PEFCs are at the mature stage, whereas those of biofuel cells (BFCs and catalysts in PEFCs are currently garnering attention. It does not mean, however, that the challenges of SOFCs and PEFC electrolytes have been overcome. SOFCs need to be operated at lower temperatures, approximately 500 °C. Electrolytes in PEFCs still suffer from a severe decrease in proton conductivity at low relative humidity and from their high cost. Catalysts in PEFCs are becoming attractive as means to reduce the platinum catalyst cost. The emerging technologies in PEFC catalysts are mainly heteroatom-doped graphene/carbon nanotubes for metal-free catalysts and supports for iron- or cobalt-based catalysts. BFCs have also received attention for wastewater treatment and as miniaturized energy sources. Of particular interest in BFCs are membrane reactors in microbial fuel cells and membrane-less enzymatic biofuel cells.

  6. LEU fuel element produced by the Egyptian fuel manufacturing pilot plant

    International Nuclear Information System (INIS)

    Zidan, W.I.

    2000-01-01

    The Egyptian Fuel Manufacturing Pilot Plant, FMPP, is a Material Testing Reactor type (MTR) fuel element facility, for producing the specified fuel elements required for the Egyptian Second Research Reactor, ETRR-2. The plant uses uranium hexafluoride (UF 6 , 19.75% U 235 by wt) as a raw material which is processed through a series of the manufacturing, inspection and test plan to produce the final specified fuel elements. Radiological safety aspects during design, construction, operation, and all reasonably accepted steps should be taken to prevent or reduce the chance of accidents occurrence. (author)

  7. Analysis of fuel cladding chemical interaction in mixed oxide fuel pins

    International Nuclear Information System (INIS)

    Weber, J.W.; Dutt, D.S.

    1976-01-01

    An analysis is presented of the observed interaction between mixed oxide 75 wt percent UO 2 --25 wt percent PuO 2 fuel and 316--20 percent CW stainless steel cladding in LMFBR type fuel pins irradiated in EBR-II. A description is given of the test pins and their operating conditions together with, metallographic observations and measurements of the fuel/cladding reaction, and a correlation equation is developed relating depth of cladding attack to temperature and burnup. Some recent data on cladding reaction in fuel pins with low initial O/M in the fuel are given and compared with the correlation equation curves

  8. 49 CFR 238.423 - Fuel tanks.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Fuel tanks. 238.423 Section 238.423 Transportation....423 Fuel tanks. (a) External fuel tanks. Each type of external fuel tank must be approved by FRA's Associate Administrator for Safety upon a showing that the fuel tank provides a level of safety at least...

  9. Review of the patents of Top and Bottom End Piece and fuel materials issued by overseas fuel vendors

    International Nuclear Information System (INIS)

    Yoon, Kyung Ho; Song, Kee Nam; Kim Hyung Kyu; Kang, Heung Seok; Bang, Je Geon.

    1997-06-01

    In order to develop new type nuclear fuel which can be loaded in the domestic PWRs without serious design changes, it is very important to understand the developing status of fuel components being conducted by overseas fuel vendors prior to commence the full-scale campaign. In this report, patents recorded in US patent lists have been inquired to understand the object, the background and the peculiarity of the invented components for top/bottom end piece and fuel materials. In addition, the important things which must be considered in development of new type fuel were described. (author). 18 tabs., 63 figs., 24 refs

  10. Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P.O. 1236909 Final report

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

    1978-12-01

    The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel

  11. Fuel economy and range estimates for fuel cell powered automobiles

    Energy Technology Data Exchange (ETDEWEB)

    Steinbugler, M.; Ogden, J. [Princeton Univ., NJ (United States)

    1996-12-31

    While a number of automotive fuel cell applications have been demonstrated, including a golf cart, buses, and a van, these systems and others that have been proposed have utilized differing configurations ranging from direct hydrogen fuel cell-only power plants to fuel cell/battery hybrids operating on reformed methanol. To date there is no clear consensus on which configuration, from among the possible combinations of fuel cell, peaking device, and fuel type, is the most likely to be successfully commercialized. System simplicity favors direct hydrogen fuel cell vehicles, but infrastructure is lacking. Infrastructure favors a system using a liquid fuel with a fuel processor, but system integration and performance issues remain. A number of studies have analyzed particular configurations on either a system or vehicle scale. The objective of this work is to estimate, within a consistent framework, fuel economies and ranges for a variety of configurations using flexible models with the goal of identifying the most promising configurations and the most important areas for further research and development.

  12. The post-irradiated examination of CANDU type fuel irradiated in the Institute for Nuclear Research TRIGA reactor

    International Nuclear Information System (INIS)

    Tuturici, I.L.; Parvan, M.; Dobrin, R.; Popov, M.; Radulescu, R.; Toma, V.

    1995-01-01

    This post-irradiation examination work has been done under the Research Contract No. 7756/RB, concluded between the International Atomic Energy Agency and the Institute for Nuclear Research. The paper contains a general description of the INR post-irradiation facility and methods and the relevant post-irradiation examination results obtained from an irradiated experimental CANDU type fuel element designed, manufactured and tested by INR in a power ramp test in the 100 kW Pressurised Water Irradiation Loop of the TRIGA 14 MW(th) Reactor. The irradiation experiment consisted in testing an assembly of six fuel elements, designed to reach a bumup of ∼ 200 MWh/kgU, with typical CANDU linear power and ramp rate. (author)

  13. Transient feedback from fuel motion in metal IFR [Integral Fast Reactor] fuel

    International Nuclear Information System (INIS)

    Rhodes, E.A.; Stanford, G.S.; Regis, J.P.; Bauer, T.H.; Dickerman, C.E.

    1990-01-01

    Results from hodoscope data analyses are presented for TREAT transient-overpower tests M5 through M7 with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding branch and prefailure elongation of D9-clad ternary (U-Pu-Zr) IFR-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT9-clad binary (U-Zr) FFTF-driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure. 4 refs., 6 figs

  14. Fuel assembly

    International Nuclear Information System (INIS)

    Yokota, Tokunobu.

    1990-01-01

    A fuel assembly used in a FBR type nuclear reactor comprises a plurality of fuel rods and a moderator guide member (water rod). A moderator exit opening/closing mechanism is formed at the upper portion of the moderator guide member for opening and closing a moderator exit. In the initial fuel charging operation cycle to the reactor, the moderator exit is closed by the moderator exit opening/closing mechanism. Then, voids are accumulated at the inner upper portion of the moderator guide member to harden spectrum and a great amount of plutonium is generated and accumulated in the fuel assembly. Further, in the fuel re-charging operation cycle, the moderator guide member is used having the moderator exit opened. In this case, voids are discharged from the moderator guide member to decrease the ratio, and the plutonium accumulated in the initial charging operation cycle is burnt. In this way, the fuel economy can be improved. (I.N.)

  15. Nondestructive analysis of irradiated fuels

    International Nuclear Information System (INIS)

    Dudey, N.D.; Frick, D.C.

    1977-01-01

    The principal nondestructive examination techniques presently used to assess the physical integrity of reactor fuels and cladding materials include gamma-scanning, profilometry, eddy current, visual inspection, rod-to-rod spacing, and neutron radiography. LWR fuels are generally examined during annual refueling outages, and are conducted underwater in the spent fuel pool. FBR fuels are primarily examined in hot cells after fuel discharge. Although the NDE techniques are identical, LWR fuel examinations emphasize tests to demonstrate adherence to technical specification and reliable fuel performance; whereas, FBR fuel examinations emphasize aspects more related to the relative performance of different types of fuel and cladding materials subjected to variable irradiation conditions

  16. Recovery of enriched Uranium (20% U-235) from wastes obtained in the preparation of fuel elements for argonaut type reactors

    International Nuclear Information System (INIS)

    Uriarte, A.; Ramos, L.; Estrada, J.; del Val, J. L.

    1962-01-01

    Results obtained with the two following installations for recovering enriched uranium (20% U-235) from wastes obtained in the preparation of fuel elements for Argonaut type reactors are presented. Ion exchange unit to recover uranium form mother liquors resulting from the precipitation ammonium diuranate (ADU) from UO 2 F 2 solutions. Uranium recovery unit from solid wastes from the process of manufacture of fuel elements, consisting of a) waste dissolution, and b) extraction with 10% (v/v) TBP. (Author) 9 refs

  17. Recovery of enriched Uranium (20% U-235) from wastes obtained in the preparation of fuel elements for argonaut type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Uriarte, A; Ramos, L; Estrada, J; Val, J L. del

    1962-07-01

    Results obtained with the two following installations for recovering enriched uranium (20% U-235) from wastes obtained in the preparation of fuel elements for Argonaut type reactors are presented. Ion exchange unit to recover uranium form mother liquors resulting from the precipitation ammonium diuranate (ADU) from UO{sub 2}F{sub 2} solutions. Uranium recovery unit from solid wastes from the process of manufacture of fuel elements, consisting of a) waste dissolution, and b) extraction with 10% (v/v) TBP. (Author) 9 refs.

  18. Nuclear fuel cycle facility accident analysis handbook

    International Nuclear Information System (INIS)

    Ayer, J.E.; Clark, A.T.; Loysen, P.; Ballinger, M.Y.; Mishima, J.; Owczarski, P.C.; Gregory, W.S.; Nichols, B.D.

    1988-05-01

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH

  19. A ECONOMIA DA EXPERIÊNCIA APLICADA A WEBSITES: uma análise dos restaurantes premiados pelo Guia Michelin na cidade de São Paulo (BRA

    Directory of Open Access Journals (Sweden)

    Pedro Henrique Drudi

    2017-09-01

    Full Text Available Este artigo tem como objetivo propor uma análise de experiência em restaurantes através de websites. Para tanto busca-se fazer uma conexão entre as esferas da experiência propostas por Pine II e Gilmore (1999 a partir da adaptação do roteiro de experiência física em restaurantes desenvolvido por Gimenes, Fraiz e Gândara (2012 para o universo virtual. A aplicação do modelo adaptado foi testada nos treze restaurantes da cidade de São Paulo contemplados com estrelas na edição de 2016 do Guia Michelin Brasil – São Paulo e Rio de Janeiro. Dentre os principais resultados verificou-se que os websites conseguem, com sucesso, materializar a proposta da casa e antecipar a experiência que será vivenciada in loco. Contudo, verifica-se algumas lacunas informativas, bem como o pouco uso, por parte de alguns websites, de recursos interativos que poderiam enriquecer a experiência de navegação.

  20. Fuel motion in overpower tests of metallic integral fast reactor fuel

    International Nuclear Information System (INIS)

    Rhodes, E.A.; Bauer, T.H.; Stanford, G.S.; Regis, J.P.; Dickerman, C.E.

    1992-01-01

    In this paper results from hodoscope data analyses are presented for transient overpower (TOP) tests M5, M6, and M7 at the Transient Reactor Test Facility, with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding breach and prefailure elongation of D9-clad ternary (U-Pu-Zr) integral fast reactor-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT-9-clad binary (U-Zr) Fast Flux Test Facility driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure

  1. Compatibility of elastomers in alternate jet fuels

    Science.gov (United States)

    Kalfayan, S. H.; Fedors, R. F.; Reilly, W. W.

    1979-01-01

    The compatibility of elastomeric compositions of known resistance to aircraft fuels was tested for potential use in Jet A type fuels obtainable from alternate sources, such as coal. Since such fuels were not available at the time, synthetic alternate fuels were prepared by adding tetralin to a petroleum based Jet A type fuel to simulate coal derived fuels which are expected to contain higher amounts of aromatic and hydroaromatic hydrocarbons. The elastomeric compounds tested were based on butadiene-acrylonitrile rubber, a castable Thiokol polysulfide rubber, and a castable fluorosilicone rubber. Batches of various cross-link densities of these rubbers were made and their chemical stress relaxation behavior in fuel, air, and nitrogen, their swelling properties, and response to mechanical testing were determined.

  2. METMET fuel with Zirconium matrix alloys

    International Nuclear Information System (INIS)

    Savchenko, A.; Konovalov, I.; Totev, T.

    2008-01-01

    The novel type of WWER-1000 fuel has been designed at A.A. Bochvar Institute. Instead of WWER-1000 UO 2 pelletized fuel rod we apply dispersion type fuel element with uniformly distributed high uranium content granules of U9Mo, U5Nb5Zr, U3Si alloys metallurgically bonded between themselves and to cladding by a specially developed Zr-base matrix alloy. The fuel meat retains a controllable porosity to accommodate fuel swelling. The optimal volume ratios between the components are: 64% fuel, 18% matrix, 18% pores. Properties of novel materials as well as fuel compositions on their base have been investigated. Method of fuel elements fabrication by capillary impregnation has been developed. The primary advantages of novel fuel are high uranium content (more than 15% in comparison with the standard UO 2 pelletized fuel rod), low temperature of fuel ( * d/tU) and serviceability under transient conditions. The use of the novel fuel might lead to natural uranium saving and reduced amounts of spent fuel as well as to optimization of Nuclear Plant operation conditions and improvements of their operation reliability and safety. As a result the economic efficiency shall increase and the cost of electric power shall decrease. (authors)

  3. Fuel Cell Power Plant Initiative. Volume 1; Solid Oxide Fuel Cell/Logistics Fuel Processor 27 kWe Power System Demonstration for ARPA

    Science.gov (United States)

    Veyo, S.E.

    1997-01-01

    This report describes the successful testing of a 27 kWe Solid Oxide Fuel Cell (SOFC) generator fueled by natural gas and/or a fuel gas produced by a brassboard logistics fuel preprocessor (LFP). The test period began on May 24, 1995 and ended on February 26, 1996 with the successful completion of all program requirements and objectives. During this time period, this power system produced 118.2 MWh of electric power. No degradation of the generator's performance was measured after 5582 accumulated hours of operation on these fuels: local natural gas - 3261 hours, jet fuel reformate gas - 766 hours, and diesel fuel reformate gas - 1555 hours. This SOFC generator was thermally cycled from full operating temperature to room temperature and back to operating temperature six times, because of failures of support system components and the occasional loss of test site power, without measurable cell degradation. Numerous outages of the LFP did not interrupt the generator's operation because the fuel control system quickly switched to local natural gas when an alarm indicated that the LFP reformate fuel supply had been interrupted. The report presents the measured electrical performance of the generator on all three fuel types and notes the small differences due to fuel type. Operational difficulties due to component failures are well documented even though they did not affect the overall excellent performance of this SOFC power generator. The final two appendices describe in detail the LFP design and the operating history of the tested brassboard LFP.

  4. Proton exchange membrane fuel cells

    CERN Document Server

    Qi, Zhigang

    2013-01-01

    Preface Proton Exchange Membrane Fuel CellsFuel CellsTypes of Fuel CellsAdvantages of Fuel CellsProton Exchange Membrane Fuel CellsMembraneCatalystCatalyst LayerGas Diffusion MediumMicroporous LayerMembrane Electrode AssemblyPlateSingle CellStackSystemCell Voltage Monitoring Module (CVM)Fuel Supply Module (FSM)Air Supply Module (ASM)Exhaust Management Module (EMM)Heat Management Module (HMM)Water Management Module (WMM)Internal Power Supply Module (IPM)Power Conditioning Module (PCM)Communications Module (COM)Controls Module (CM)SummaryThermodynamics and KineticsTheoretical EfficiencyVoltagePo

  5. Liquid fuel cells

    Directory of Open Access Journals (Sweden)

    Grigorii L. Soloveichik

    2014-08-01

    Full Text Available The advantages of liquid fuel cells (LFCs over conventional hydrogen–oxygen fuel cells include a higher theoretical energy density and efficiency, a more convenient handling of the streams, and enhanced safety. This review focuses on the use of different types of organic fuels as an anode material for LFCs. An overview of the current state of the art and recent trends in the development of LFC and the challenges of their practical implementation are presented.

  6. Thermodynamic analysis of carbon formation in solid oxide fuel cells with a direct internal reformer fueled by ethanol, methanol, and methane

    International Nuclear Information System (INIS)

    Laosiripojana, N.; Assabumrungrat, S.; Pavarajarn, V.; Sangtongkitcharoen, W.; Tangjitmatee, A.; Praserthdam, P.

    2004-01-01

    'Full text:' This paper concerns a detailed thermodynamic analysis of carbon formation for a Direct Internal Reformer (DIR) Solid Oxide Fuel Cells (SOFC). The modeling of DIR-SOFC fueled by ethanol, methanol, and methane were compared. Two types of fuel cell electrolytes, i.e. oxygen-conducting and hydrogen-conducting, are considered. Equilibrium calculations were performed to find the ranges of inlet steam/fuel ratio where carbon formation is thermodynamically unfavorable in the temperature range of 500-1200 K. It was found that the key parameters determining the boundary of carbon formation are temperature, type of solid electrolyte and extent of the electrochemical reaction of hydrogen. The minimum requirements of H2O/fuel ratio for each type of fuel in which the carbon formation is thermodynamically unfavored were compared. At the same operating conditions, DIR-SOFC fueled by ethanol required the lowest inlet H2O/fuel ratio in which the carbon formation is thermodynamically unfavored. The requirement decreased with increasing temperature for all three fuels. Comparison between two types of the electrolytes reveals that the hydrogen-conducting electrolyte is impractical for use, regarding to the tendency of carbon formation. This is due mainly to the water formed by the electrochemical reaction at the electrodes. (author)

  7. Transparency associated with the nuclear fuel cycle

    International Nuclear Information System (INIS)

    2009-01-01

    This document presents the different fuel cycle stages with which the CEA is associated, the annual flow of materials and wastes produced at these different stages, and the destiny of these produced materials and wastes. These information are given for the different CEA R and D activities: experimentation hot laboratories (activities, fuel cycle stages, list of laboratories, tables giving annual flows for each of them), research reactors (types of reactors, fuel usage modes, annual flows of nuclear materials for each reactor), spent fuel management (different types of used materials), spent fuels and radioactive wastes with a foreign origin (quantities, processes)

  8. Nuclear fuels for material test reactors

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Durazzo, M.; Freitas, C.T. de

    1982-01-01

    Experimental results related do the development of nuclear fuels for reactors cooled and moderated by water have been presented cylindrical and plate type fuels have been described in which the core consists of U compouns dispersed in an Al matrix and is clad with aluminium. Fabrication details involving rollmilling, swaging or hot pressing have been described. Corrosion and irradiation test results are also discussed. The performance of the different types of fuels indicates that it is possible to locally fabricate fuel plates with U 3 O 8 +Al cores (20% enriched U) for use in operating Brazilian research reactors. (Author) [pt

  9. Postirradiation Examination Of U3O8-AL Plate Type Dispersion Fuel Element

    International Nuclear Information System (INIS)

    Nasution-Hasbullah; Sugondo; Amin, D.L.; Siti-Amini

    1996-01-01

    Postirradiation examination of plate type spent fuel element RIE-01 has been carried out in order to observer its physical changes and performance under irradiation in the reactor. The irradiation has been time more than two years with a declared burnup of 51.04 %. The examination included visual and dimensional measurement, measurement of burn-up distribution, wipe test and metallographic analysis. The results showed that all fuel plates retained their integrity. The colour changes were occurred on most of the plates significant suggesting that it was generated from the oxide layer formation. From gamma-scanning examination it could be deducted that the highest burn-up distribution of the plate was at position of 30 cm from the bottom. A more homogeneous distribution was found in the middle plate of the bundle. The increased plate thickness, as revealed by dimensional measurements as in agreement with the burn-up distribution pattern. Despite the changes observed in could be concluded that all changes occurred were still within the allowable limits and therefore it can recommended that an increase of the burn-up level above 51,04 % is still quite possible

  10. Biological fuel cells and their applications

    OpenAIRE

    Shukla, AK; Suresh, P; Berchmans, S; Rajendran, A

    2004-01-01

    One type of genuine fuel cell that does hold promise in the long-term is the biological fuel cell. Unlike conventional fuel cells, which employ hydrogen, ethanol and methanol as fuel, biological fuel cells use organic products produced by metabolic processes or use organic electron donors utilized in the growth processes as fuels for current generation. A distinctive feature of biological fuel cells is that the electrode reactions are controlled by biocatalysts, i.e. the biological redox-reac...

  11. Gel-sphere-pac reactor fuel fabrication and its application to a variety of fuels

    International Nuclear Information System (INIS)

    Olsen, A.R.; Judkins, R.R.

    1979-12-01

    The gel-sphere-pac fuel fabrication option was evaluated for its possible application to commercial scale fuel fabrication for 19 fuel element designs that use oxide fuel in metal clad rods. The dry gel spheres are prepared at the reprocessing plant and are then calcined, sintered, inspected, and loaded into fuel rods and packed by low-energy vibration. A fuel smear density of 83 to 88% theoretical can be obtained. All fuel fabrication process steps were defined and evaluated from fuel receiving to finished fuel element shipping. The evaluation also covers the feasibility of the process, the current status of technology, estimates of the required time and cost to develop the technology to commercial status, and the safety and licensability of commercial scale plants. The primary evaluation was for a Light-Water Reactor fuel element containing (U,Pu)O 2 fuel. The other 18 fuel element types - 3 for Light-Water Reactors, 1 for a Heavy-Water Reactor, 1 for a Gas-Cooled Fast Reactor, 7 for Liquid-Metal-Cooled Fast Breeder Reactors, and 3 pairs for Light-Water Prebreeder and Breeder Reactors - were compared with the Light-Water Reactor. The gel-sphere-pac option was found applicable to 17 of the 19 element types; the characteristics of a commercial scale plant were defined for these for making cost estimates for such plants. The evaluation clearly shows the gel-sphere-pac process to be a viable fuel fabrication option. Estimates indicate a significant potential fabrication cost advantage for the gel-sphere-pac process if a remotely operated and remotely maintained fuel fabrication plant is required

  12. Fuel element structure - design, production and operational behaviour

    International Nuclear Information System (INIS)

    Pott, G.; Dietz, W.

    1985-01-01

    The lectures held at the meeting of the fuel element section of the Kerntechnische Gesellschaft gives a survey of developments in fuel element structure design for PWR-type, BWR-type and fast breeder reactors. For better utilization of the fuel, concepts have been developed for re-usable, removable and thus repairable fuel elements. Furthermore, the manufacturing methods for fuel element structures were refined to achieve better quality and more efficient manufacturing methods. Statements on the dimensional behaviour and on the mechanical stability of fuel element structures in normal and accident operation could be made on the basis of post-irradiation inspections. Finally, the design, manufacture and irradiation behaviour of graphite reflectors in HTGR-type reactors are described. The 12 lectures have been recorded in the data base separately. (RF) [de

  13. Fuel cells

    International Nuclear Information System (INIS)

    Niederdoeckl, J.

    2001-01-01

    Europe has at present big hopes on the fuel cells technology, in comparison with other energy conversion technologies, this technology has important advantages, for example: high efficiency, very low pollution and parallel use of electric and thermal energy. Preliminary works for fuel cells developing and its commercial exploitation are at full speed; until now the European Union has invested approx. 1.7 billion Schillings, 60 relevant projects are being executed. The Austrian industry is interested in applying this technique to drives, thermal power stations and the miniature fuel cells as replacement of batteries in electronic products (Notebooks, mobile telephones, etc.). A general description of the historic development of fuel cells including the main types is given as well as what is the situation in Austria. (nevyjel)

  14. Natural uranium metallic fuel elements: fabrication and operating experience

    International Nuclear Information System (INIS)

    Hammad, F.H.; Abou-Zahra, A.A.; Sharkawy, S.W.

    1980-01-01

    The main reactor types based on natural uranium metallic fuel element, particularly the early types, are reviewed in this report. The reactor types are: graphite moderated air cooled, graphite moderated gas cooled and heavy water moderated reactors. The design features, fabrication technology of these reactor fuel elements and the operating experience gained during reactor operation are described and discussed. The interrelation between operating experience, fuel design and fabrication was also discussed with emphasis on improving fuel performance. (author)

  15. Behaviour of rock-like oxide fuels under reactivity-initiated accident conditions

    International Nuclear Information System (INIS)

    Kazuyuki, Kusagaya; Takehiko, Nakamura; Makio, Yoshinaga; Hiroshi, Akie; Toshiyuki, Yamashita; Hiroshi, Uetsuka

    2002-01-01

    Pulse irradiation tests of three types of un-irradiated rock-like oxide (ROX) fuel - yttria-stabilised zirconia (YSZ) single phase, YSZ and spinel (MgAl 2 O 4 ) homogeneous mixture and particle-dispersed YSZ/spinel - were conducted in the Nuclear Safety Research Reactor to investigate the fuel behaviour under reactivity-initiated accident conditions. The ROX fuels failed at fuel volumetric enthalpies above 10 GJ/m 3 , which was comparable to that of un-irradiated UO 2 fuel. The failure mode of the ROX fuels, however, was quite different from that of the UO 2 fuel. The ROX fuels failed with fuel pellet melting and a part of the molten fuel was released out to the surrounding coolant water. In spite of the release, no significant mechanical energy generation due to fuel/coolant thermal interaction was observed in the tested enthalpy range below∼12 GJ/m 3 . The YSZ type and homogenous YSZ/spinel type ROX fuels failed by cladding burst when their temperatures peaked, while the particle-dispersed YSZ/spinel type ROX fuel seemed to have failed by cladding local melting. (author)

  16. Review of oxidation rates of DOE spent nuclear fuel : Part 1 : nuclear fuel

    International Nuclear Information System (INIS)

    Hilton, B.A.

    2000-01-01

    The long-term performance of Department of Energy (DOE) spent nuclear fuel (SNF) in a mined geologic disposal system depends highly on fuel oxidation and subsequent radionuclide release. The oxidation rates of nuclear fuels are reviewed in this two-volume report to provide a baseline for comparison with release rate data and technical rationale for predicting general corrosion behavior of DOE SNF. The oxidation rates of nuclear fuels in the DOE SNF inventory were organized according to metallic, Part 1, and non-metallic, Part 2, spent nuclear fuels. This Part 1 of the report reviews the oxidation behavior of three fuel types prototypic of metallic fuel in the DOE SNF inventory: uranium metal, uranium alloys and aluminum-based dispersion fuels. The oxidation rates of these fuels were evaluated in oxygen, water vapor, and water. The water data were limited to pure water corrosion as this represents baseline corrosion kinetics. Since the oxidation processes and kinetics discussed in this report are limited to pure water, they are not directly applicable to corrosion rates of SNF in water chemistry that is significantly different (such as may occur in the repository). Linear kinetics adequately described the oxidation rates of metallic fuels in long-term corrosion. Temperature dependent oxidation rates were determined by linear regression analysis of the literature data. As expected the reaction rates of metallic fuels dramatically increase with temperature. The uranium metal and metal alloys have stronger temperature dependence than the aluminum dispersion fuels. The uranium metal/water reaction exhibited the highest oxidation rate of the metallic fuel types and environments that were reviewed. Consequently, the corrosion properties of all DOE SNF may be conservatively modeled as uranium metal, which is representative of spent N-Reactor fuel. The reaction rate in anoxic, saturated water vapor was essentially the same as the water reaction rate. The long-term intrinsic

  17. MOX fuel assembly and reactor core

    International Nuclear Information System (INIS)

    Shimada, Hidemitsu; Koyama, Jun-ichi; Aoyama, Motoo

    1998-01-01

    The MOX fuel assembly of the present invention is of a c-lattice type loaded to a BWR type reactor. 74 MOX fuel rods filled with mixed oxides of uranium and plutonium and two water rods disposed to a space equal to that for 7 MOX fuel rods are arranged in 9 x 9 matrix. MOX fuel rods having the lowest enrichment degree are disposed to four corners of the 9 x 9 matrix. The enrichment degree means a ratio of the weight of fission products based on the total weight of fuels. Two MOX fuel rods having the same enrichment degree are arranged in each direction so as to be continuous from the MOX fuel rods at four corners in the direction of the same row and different column and same column and the different row. In addition, among the outermost circumferential portion of the 9 x 9 matrix, MOX fuel rods having a lower enrichment degree next to the MOX fuel rods having the lowest enrichment degree are arranged, each by three to a portion where MOX fuel rods having the lowest enrichment degree are not disposed. (I.N.)

  18. Results of experiments with flare type igniters on diesel fuel and crude oil emulsions

    International Nuclear Information System (INIS)

    Moffat, C.; Hankins, P.

    1997-01-01

    Development of a hand-deployable igniter that could ignite contained diesel fuel and crude oil emulsions on water was described. The igniter was developed as part of the U.S. Navy Supervisor of Salvage (SUPSALV) In-Situ Burn (ISB) system. It is a manually operated, electrically fired, high temperature flare type igniter. It is 41 cm long, 10 cm in diameter, weighs 1.5 kg, and is packaged and shipped with the ISB system. The chemical and mineral composition of the flair allows for a three minute burn of up to 1370 degrees C (2500 degrees F) at the center. The flare is most effective when used in conjunction with a shroud of sorbent material which traps and holds oil around the burning flare aiding the ignition process by increasing the initial propagation area. In small-scale tank experiments the flare ignited diesel fuel in ambient temperatures of 3 degrees C, with winds of 8 to 10 m/sec. The flare also ignited 22.5 per cent water-in crude oil emulsion in 3 degrees C temperatures. 4 refs., 3 tabs

  19. Fuel cycle in Japanese Fugen - HWR

    International Nuclear Information System (INIS)

    1979-04-01

    This paper describes the use of plutonium-bearing fuel in the Japanese Fugen-HWR. The Fugen-HWR is a pressure tube type, boiling light water cooled, and heavy water moderated reactor, which by using plutonium fuel (MOX) achieves the advantage of high neutron economy. The characteristics of the reactor are discussed, particularly its ability to operate with several different types of fuel - Pu-natural U MOX, Pu-Depleted U (from spent LWR fuel) MOX, Pu-Depleted U (from enrichment tails) MOX, and enriched UO 2 . The natural U and separative work units saved are given and the fuel management and control of the reactor discussed. Non-proliferation and safety considerations are given. The Fugen-HWR achieved 100% power rating in the autumn of 1979

  20. Trapped in the heat: A post-communist type of fuel poverty

    International Nuclear Information System (INIS)

    Tirado Herrero, Sergio; Ürge-Vorsatz, Diana

    2012-01-01

    Fuel poverty is a still insufficiently researched social and energy challenge with significant climate change implications. Based on evidence from Hungarian panel apartment blocks connected to district heating, this paper introduces a new variant of fuel poverty that may not be properly captured by existing fuel poverty indicators. This newly defined variant can be largely attributed to post-communist legacies – though it might also exist in other contexts – and assumes that consumers living in poor-efficiency, district-heated buildings are trapped in dwellings with adequate indoor temperatures but disproportionately high heating costs because (a) changing supplier or fuel is difficult because of the existing technical and institutional constraints, and (b) they do not realistically have the option to reduce individually their heating costs through individual efficiency improvements. This situation often translates into payment arrears, indebtedness, risk of disconnection, or reduced consumption of other basic goods and services. State-supported policy responses to date have favoured symptomatic solutions (direct consumer support) combined with superficial retrofits, though it is argued that only state-of-the-art retrofits such as the passive house-based SOLANOVA pilot project in Dunaújváros can fully eradicate fuel poverty in this consumer group. - Highlights: ► We identify a new variant of fuel poverty. ► We explore this variant in panel apartment blocks connected to DH in Hungary, where dwellings are warm enough in winter but have disproportionately high energy costs. ► Affected households react in ways that harm their welfare and put them at risk. ► Deep retrofits in dwellings such as these can eradicate fuel poverty while also contributing to other goals.

  1. Fuel Performance Modeling of U-Mo Dispersion Fuel: The thermal conductivity of the interaction layers of the irradiated U-Mo dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mistarhi, Qusai M.; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    U-Mo/Al dispersion fuel performed well at a low burn-up. However, higher burn-up and higher fission rate irradiation testing showed enhanced fuel meat swelling which was caused by high interaction layer growth and pore formation. The performance of the dispersion type fuel in the irradiation and un-irradiation environment is very important. During the fabrication of the dispersion type fuel an Interaction Layer (IL) is formed due to the inter-diffusion between the U-Mo fuel particles and the Al matrix which is an intermetallic compound (U,Mo)Alx. During irradiation, the IL becomes amorphous causing a further decrease in the thermal conductivity and an increase in the centerline temperature of the fuel meat. Several analytical models and numerical methods were developed to study the performance of the unirradiated U-Mo/Al dispersion fuel. Two analytical models were developed to study the performance of the irradiated U-Mo/Al dispersion fuel. In these models, the thermal conductivity of the IL was assumed to be constant. The properties of the irradiated U-Mo dispersion fuel have been investigated recently by Huber et al. The objective of this study is to develop a correlation for IL thermal conductivity during irradiation as a function of the temperature and fission density from the experimentally measured thermal conductivity of the irradiated U-Mo/Al dispersion fuel. The thermal conductivity of IL during irradiation was calculated from the experimentally measured data and a correlation was developed from the thermal conductivity of IL as a function of T and fission density.

  2. Fuel assembly, channel box of fuel assembly, fuel spacer of fuel assembly and method of manufacturing channel box

    International Nuclear Information System (INIS)

    Chaki, Masao; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Nishida, Koji; Kawasaki, Terufumi.

    1997-01-01

    In a fuel assembly of a BWR type reactor, fuel rods disposed at corners of side walls of a channel box or in the periphery of the side walls are partially removed, and recessed portions are formed on the side walls of the channel box from which the fuel rods are removed. Spaces closed at the sides are formed in the inner side of the corner portions. Openings are formed for communicating the closed space with the outside of the channel box. Then, the channel area of the outer side of the channel box is increased, through which much water flows to increase the amount of water in the reactor core thereby promoting the moderation of neutrons and providing thermal neutrons suitable to nuclear fission. The degree of freedom for distribution of the spaces in the reactor core is increased to improve neutron economy thereby enabling to utilize reactor fuels effectively. (N.H.)

  3. 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

  4. IFBA credit in the Shearon Harris fuel racks with Vantage 5 fuel

    International Nuclear Information System (INIS)

    Boyd, W.A.; Schmidt, R.F.; Erwin, R.D.

    1989-01-01

    At the Shearon Harris nuclear plant, fuel management strategies are being considered which will result in feed fuel enrichments approaching 5.0 w/o U-235. These types of enrichments require a new criticality analysis to raise the existing fuel rack enrichment limit. It is receiving Westinghouse Vantage 5 fuel with integral fuel burnable absorber (IFBA) rods providing the depletable neutron absorber. An analysis was performed on the fuel racks which demonstrates that fuel enriched up to 5.0 w/o U-235 can be stored by taking credit for the IFBA rods present in the high enriched fuel assemblies. This is done by calculating the maximum Vantage 5 fuel assembly reactivity that can be placed in the fuel racks and meet the criticality K-eff limit. A methodology is also developed which conservatively calculates the minimum number of IFBA rods needed per assembly to meet the fuel rack storage limits. This eliminates the need for core designers to determine assembly K-inf terms for every different enrichment/IFBA combination

  5. Evaluation of Electron Beam Welding Performance of AA6061-T6 Plate-type Fuel Assembly

    International Nuclear Information System (INIS)

    Kim, Soo-Sung; Seo, Kyoung-Seok; Lee, Don-Bae; Park, Jong-Man; Lee, Yoon-Sang; Lee, Chong-Tak

    2014-01-01

    As one of the most commonly used heat-treatable aluminum alloys, AA6061-T6 aluminum alloy is available in a wide range of structural materials. Typically, it is used in structural members, auto-body sheet and many other applications. Generally, this alloy is easily welded by conventional GTAW (Gas Tungsten Arc Welding), LBW (Laser Beam Welding) and EBW(Electron Beam Welding). However, certain characteristics, such as solidification cracking, porosity, HAZ (Heat-affected Zone) degradation must be considered during welding. Because of high energy density and low heat input, especially LBW and EBW processes possess the advantage of minimizing the fusing zone and HAZ and producing deeper penetration than arc welding processes. In present study, to apply for the plate-type nuclear fuel fabrication and assembly, a fundamental electron beam welding experiment using AA6061-T6 aluminum alloy specimens was conducted. Furthermore, to establish the suitable welding process, and satisfy the requirements of the weld quality, EBW apparatus using an electron welding gun and vacuum chamber was developed, and preliminary investigations for optimizing the welding parameters of the specimens using AA6061-T6 aluminum plates were also performed. The EB weld quality of AA6061-T6 aluminum alloy for the plate-type fuel assembly has been also studied by the weld penetrations of side plate to end fitting and fixing bar and weld inspections using computed tomography

  6. Fuel assembly

    International Nuclear Information System (INIS)

    Kawai, Mitsuo.

    1988-01-01

    Purpose: To reduce the corrosion rate and suppress the increase of radioactive corrosion products in reactor water of nuclear fuel assemblies for use in BWR type reactors having spacer springs made of nickel based deposition reinforced type alloys. Constitution: Spacer rings made of nickel based deposition reinforced type alloy are incorporated and used as fuel assemblies after applying treatment of dipping and maintaining at high temperature water followed by heating in steams. Since this can remove the nickel leaching into reactor water at the initial stage, Co-58 as the radioactive corrosion products in the reactor water can be reduced, and the operation at in-service inspection or repairement can be facilitated to improve the working efficiency of the nuclear power plant. The dipping time is desirably more than 10 hours and more desirably more than 30 hours. (Horiuchi, T. )

  7. Proceedings of fuel strategies for conventional and unconventional fuels

    International Nuclear Information System (INIS)

    Mahr, D.; Nechvatal, T.T.

    1991-01-01

    Fuel selection is a major decision for a power engineer. It is the single largest item in the power plant operating budget and has a major effect on power plant economics. Fuel determines plant design requirements and the types of systems that are provided. As a result, it affects capital budgets and financing requirements. In the last few decades, we have seen different fuels of choice at any one time. Coal has always been a staple for power generation. During the 1950s and 1960s, oil became an attractive alternative. Nuclear fuel became a popular choice due to its very low energy cost. After Three-Mile-Island, however, capital budgets went through the roof, resulting in severe financial constraints. Natural gas, which was rationed in some regions a few years ago, is now a popular choice. Some sources predict that its cost will increase faster than other fuels. To mitigate the relative variations in energy cost for different fuels, a balanced energy plan is required. A balanced power generation plan with fuel options provides the flexibility to react to unpredictable changes. The papers in this book are a continuation of the Fuel Strategies theme. Three technical topics are covered: Converting to Orimulsion, A Replacement Fuel for Heavy Oil; Innovations in Handling Conventional and Unconventional Fuels for Power Plants; and Pacific Rim Experience With Coal

  8. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part)

    International Nuclear Information System (INIS)

    Hernandez L, H.; Ortiz V, J.

    2003-01-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  9. Comparison of fuel production costs for future transportation

    DEFF Research Database (Denmark)

    Ridjan, Iva; Mathiesen, Brian Vad; Connolly, David

    The purpose of this poster is to provide an overview of fuel production costs for two types of synthetic fuels – methanol and methane, along with comparable costs for first and second generation biodiesel, two types of second generation bioethanol, and biogas. The model analysed is a 100% renewable...... scenario of Denmark for 2050, where the data for the transport sector has been changed to estimate the fuel production costs for eight different fuel pathways....

  10. Pressure drop calculation in a fuel element of a pool type reactor

    International Nuclear Information System (INIS)

    Lassance, Victor; Oliveira, Andre F.; Moreira, Maria de L.

    2013-01-01

    Even with the advances of hardware in computer sciences, sometimes it is necessary to simplify the simulation in order to optimize the results given the same calculation runtime. The object of this study is a thermodynamic analysis of the core of a pool type research reactor, focusing on natural circulation. Due to the high geometrical complexity of the core, the scale transfer process becomes an essential step to the thermodynamic study of the reactor. This process takes place by determining the effective equivalent properties obtained from a detailed simulation of the core and transferring them to a porous medium having a coarse mesh while preserving the overall characteristics. In this way, it will be able to obtain the quadratic resistance coefficient KQ by calculating the pressure drop inside the fuel element. To observe in detail the behavior of this flow, longitudinal and transversal cross sections will be made in different points, thereby observing the velocity and pressure distributions. The analysis will provide detailed data on the fluid flow between the fuel plates enabling the observation of possible critical points or undesired behavior. The whole analysis was made by using the commercial code ANSYS CFX ver. 12.1. This is study will provide data, as a first step to enable future simulations which will consider the entire reactor. (author)

  11. Fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Akiyoshi; Bessho, Yasunori; Aoyama, Motoo; Koyama, Jun-ichi; Hirakawa, Hiromasa; Yamashita, Jun-ichi; Hayashi, Tatsuo

    1998-01-01

    In a fuel assembly of a BWR type reactor in which a water rod of a large diameter is disposed at the central portion, the cross sectional area perpendicular to the axial direction comprises a region a of a fuel rod group facing to a wide gap water region to which a control rod is inserted, a region b of a fuel rod group disposed on the side of the wide gap water region other than the region a, a region d of a fuel rod group facing to a narrow gap water region and a region c of a fuel rod group disposed on the side of the narrow gap water region other than the region d. When comparing an amount of fission products contained in the four regions relative to that in the entire regions and average enrichment degrees of fuel rods for the four regions, the relative amount and the average enrichment degree of the fuel rod group of the region a is minimized, and the relative amount and the average enrichment degree of the fuel rod group in the region b is maximized. Then, reactor shut down margin during cold operation can be improved while flattening the power in the cross section perpendicular to the axial direction. (N.H.)

  12. CASTOR(r) and CONSTOR(r) type transport and storage casks for spent fuel and high active waste

    International Nuclear Information System (INIS)

    Kuehne, B.; Sowa, W.

    2002-01-01

    The German company GNB has developed, tested, licensed, fabricated, loaded, transported and stored a large number of casks for spent fuel and high-level waste. CASTOR(r) casks are used at 18 sites on three continents. Spent fuel assemblies of the types PWR, BWR, VVER, RBMK, MTR and THTR as well as vitrified high active waste (HAW) containers are stored in these kinds of casks. More than 600 CASTOR(r) casks have been loaded for long-term storage. The two decades of storage have shown that the basic requirements, which are safe confinement, criticality safety, sufficient shielding and appropriate heat transfer have been fulfilled in each case. There is no indication that problems will arise in the future. Of course, the experience of 20 years has resulted in improvements of the cask design. One basic improvement is GNB's development since the mid 1990s of a sandwich cask design using heavy concrete and steel as basic materials, for economical and technical reasons. This CONSTOR(r) cask concept also fulfils all design criteria for transport and storage given by the IAEA recommendations and national authorities. By May 2002 40 CONSTOR(r) casks had been delivered and 15 had been successfully loaded and stored. In this paper the different types of casks are presented. Experiences gained during the large number of cask loadings and more than 4000 cask-years of storage will be summarised. The presentation of recent and future development shows the optimisation potential of the CASTOR(r) and CONSTOR(r) cask families for safe and economical management of spent fuel. (author)

  13. A cidade desejada e sublimada por Jorge Amado: os lugares imaginados em Bahia de Todos-os- Santos: guia de ruas e mistérios de Salvador

    Directory of Open Access Journals (Sweden)

    Ricardo Araújo Barberena

    2013-01-01

    Full Text Available The book Bahia de Todos-os-santos: guia de ruas e mistérios de Salvador (Bahia of all-saints: a guide to the streets and mysteries of Salvador, by Jorge Amado (published in 1944, portrays an easygoing and provincial capital of Bahia. The population, which was less than 300 thousand people, moved about in the different celebrations of life. Due to profound urban transformations, the book underwent alterations in the different versions published through the years. However, the underlying structure and the spirit of the book have remained: the production of an encyclopedia of what it means to be/being “baiano” – sceneries, stories, old streets, new avenues, traditions, parties, poverty, joy, churches, candomblé, orishas, and other characters. In its pages, the book presents a real and magical image of a territory permeated by ordinary mysteries. The book has maintained its essence throughout the different editions. Even if the city has changed physically, it remains unchanged in terms of its poetic prose and in the production of a sublime descriptivism of a “black Rome.”

  14. Fuel cells for commercial energy

    Science.gov (United States)

    Huppmann, Gerhard; Weisse, Eckart; Bischoff, Manfred

    1990-04-01

    The development of various types of fuel cells is described. Advantges and drawbacks are considered for alkaline fuel cells, phosphoric acid fuel cells, and molten carbonate fuel cells. It is shown that their modular construction is particularly adapted to power heat systems. A comparison which is largely in favor of fuel cells, is made between coal, oil, natural gas power stations, and fuel cells. Safety risks in operation are also compared with those of conventional power stations. Fuel cells are particularly suited for dwellings, shopping centers, swimming pools, other sporting installations, and research facilities, whose high current and heat requirements can be covered by power heat coupling.

  15. Handbook on process and chemistry on nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Atsuyuki [Tokyo Univ., Tokyo (Japan); Asakura, Toshihide; Adachi, Takeo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; and others

    2001-12-01

    'Wet-type' nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of 'wet-type' reprocessing, because it contributes to establish and develop fuel reprocessing process and nuclear fuel cycle treating high burn-up UO{sub 2} fuel and spent MOX fuel, and to utilize 'wet-type' reprocessing technology much widely. This handbook summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing', from FY 1993 until FY 2000. (author)

  16. Handbook on process and chemistry on nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Suzuki, Atsuyuki; Asakura, Toshihide; Adachi, Takeo

    2001-12-01

    'Wet-type' nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of 'wet-type' reprocessing, because it contributes to establish and develop fuel reprocessing process and nuclear fuel cycle treating high burn-up UO 2 fuel and spent MOX fuel, and to utilize 'wet-type' reprocessing technology much widely. This handbook summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing', from FY 1993 until FY 2000. (author)

  17. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

    Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

  18. Alternative fossil-based transportation fuels

    Science.gov (United States)

    2008-01-01

    "Alternative fuels derived from oil sands and from coal liquefaction can cost-effectively diversify fuel supplies, but neither type significantly reduces U.S. carbon-dioxide emissions enough to arrest long-term climate change".

  19. Alternatives to traditional transportation fuels 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-01

    This report provides information on transportation fuels other than gasoline and diesel, and the vehicles that use these fuels. The Energy Information Administration (EIA) provides this information to support the U.S. Department of Energy`s reporting obligations under Section 503 of the Energy Policy Act of 1992 (EPACT). The principal information contained in this report includes historical and year-ahead estimates of the following: (1) the number and type of alterative-fueled vehicles (AFV`s) in use; (2) the consumption of alternative transportation fuels and {open_quotes}replacement fuels{close_quotes}; and (3) the number and type of alterative-fueled vehicles made available in the current and following years. In addition, the report contains some material on special topics. The appendices include a discussion of the methodology used to develop the estimates (Appendix A), a map defining geographic regions used, and a list of AFV suppliers.

  20. Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

    Energy Technology Data Exchange (ETDEWEB)

    Heidet, F.; Kim, T.; Grandy, C. (Nuclear Engineering Division)

    2012-07-30

    Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium is more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to determine the

  1. TRIGA low enrichment fuel

    International Nuclear Information System (INIS)

    Gietzen, A.

    1993-01-01

    Sixty TRIGA reactors have been sold and the earliest of these are now passing twenty years of operation. All of these reactors use the uranium zirconium hydride fuel (UZrH) which provides certain unique advantages arising out of its large prompt negative temperature coefficient, very low fission product release, and high temperature capability. Eleven of these Sixty reactors are conversions from plate fuel to TRIGA fuel which were made as a result of these advantages. With only a few exceptions, TRIGA reactors have always used low-enriched uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on nonproliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U.S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of 1978, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with General Atomic's standard commercial warranty

  2. TRIGA low enrichment fuel

    International Nuclear Information System (INIS)

    Gietzen, A.

    1993-01-01

    Sixty TRIGA reactors have been sold and the earliest of these are now passing twenty years of operation. All of these reactors use the uranium-zirconium hydride fuel (UZrH) which provides certain unique advantages arising out of its large prompt negative temperature coefficient, very low fission product release, and high temperature capability. Eleven of these Sixty reactors are conversions from plate fuel to TRIGA fuel which were made as a result of these advantages. With only a few exceptions, TRIGA reactors have always used low-enriched-uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on nonproliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U.S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of 1978, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with GA's standard commercial warranty

  3. Fuel assemblies

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi.

    1979-01-01

    Purpose: To prevent scattering of gaseous fission products released from fuel assemblies stored in an fbr type reactor. Constitution; A cap provided with means capable of storing gas is adapted to amount to the assembly handling head, for example, by way of threading in a storage rack of spent fuel assemblies consisting of a bottom plate, a top plate and an assembly support mechanism. By previously eliminating the gas inside of the assembly and the cap in the storage rack, gaseous fission products upon loading, if released from fuel rods during storage, are stored in the cap and do not scatter in the storage rack. (Horiuchi, T.)

  4. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B{sub 4}C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B{sub 4}C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B{sub 4}C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H{sub 2} generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B{sub 4}C were not treated. In general the confidence in code predictions decreases with processing core damage. 36 refs.

  5. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    International Nuclear Information System (INIS)

    1996-01-01

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B 4 C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B 4 C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B 4 C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H 2 generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B 4 C were not treated. In general the confidence in code predictions decreases with processing core damage. (N.T.)

  6. Studies and research concerning BNFP: spent fuel dry storage studies at the Barnwell Nuclear Fuel Plant

    International Nuclear Information System (INIS)

    Anderson, K.J.

    1980-09-01

    Conceptual designs are presented utilizing the Barnwell Nuclear Fuel Plant for the dry interim storage of spent light water reactor fuel. Studies were conducted to determine feasible approaches to storing spent fuel by methods other than wet pool storage. Fuel that has had an opportunity to cool for several years, or more, after discharge from a reactor is especially adaptable to dry storage since its thermal load is greatly reduced compared to the thermal load immediately following discharge. A thermal analysis was performed to help in determining the feasibility of various spent fuel dry storage concepts. Methods to reject the heat from dry storage are briefly discussed, which include both active and passive cooling systems. The storage modes reviewed include above and below ground caisson-type storage facilities and numerous variations of vault, or hot cell-type, storage facilities

  7. Fire behavior in masticated fuels: a review

    Science.gov (United States)

    Jesse K. Kreye; Nolan W. Brewer; Penelope Morgan; J. Morgan Varner; Alistair M.S. Smith; Chad M. Hoffman; Roger D. Ottmar

    2014-01-01

    Mastication is an increasingly common fuels treatment that redistributes ‘‘ladder’’ fuels to the forest floor to reduce vertical fuel continuity, crown fire potential, and fireline intensity, but fuel models do not exist for predicting fire behavior in these fuel types. Recent fires burning in masticated fuels have behaved in unexpected and contradictory ways, likely...

  8. Thermal neutron spectrum distribution in TRIGA fuels

    International Nuclear Information System (INIS)

    Gui Ah Auu; Harasawa, Susumu; An, Shigehiro

    1989-01-01

    The dependence of thermal neutron spectrum in TRIGA fuel cell on fuel temperature and TRIGA fuel types were studied using LIBP and THERMOS codes. Some characteristics of the TRIGA fuel including its prompt negative temperature coefficient of reactivity were explained using the results of the study. (author)

  9. ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

    Energy Technology Data Exchange (ETDEWEB)

    Skutnik, Steven E. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering

    2017-06-19

    processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.

  10. Spent fuel management in Canada

    International Nuclear Information System (INIS)

    Khan, A.; Pattantyus, P.

    1999-01-01

    The current status of the Canadian spent fuel storage is presented. This includes wet and dry interim storage. Extension of wet interim storage facilities is nor planned, as dry technologies have found wide acceptance. The Canadian nuclear program is sustained by commercial Ontario Hydro CANDU type reactors, since 1971, representing 13600 MW(e) of installed capacity, able to produce 9200 spent fuel bundles (1800 tU) every year, and Hydro Quebec and New Brunswick CANDU reactors each producing 685 MW(e) and about 100 tU of spent fuel annually. The implementation of various interim (wt and dry) storage technologies resulted in simple, dense and low cost systems. Economical factors determined that the open cycle option be adopted for the CANDU type reactors rather that recycling the spent fuel. Research and development activities for immobilization and final disposal of nuclear waste are being undertaken in the Canadian Nuclear Fuel Waste Management Program

  11. Qualification status of LEU [low enriched uranium] fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.

    1987-01-01

    Sufficient data has been obtained from tests of high-density, low-enriched fuels for research and test reactors to declare them qualified for use. These fuels include UZrH x (TRIGA fuel) and UO 2 (SPERT fuel) for rod-type reactors and UAl x , U 3 O 8 , U 3 Si 2 , and U 3 Si dispersed in aluminium for plate-type reactors. Except for U 3 Si, the allowable fission density for LEU applications is limited only by the available 235 U. Several reactors are now using these fuels, and additional conversions are in progress. The basic performance characteristics and limits, if any, of the qualified low-enriched (and medium-enriched) fuels are discussed. Continuing and planned work to qualify additional fuels is also discussed. (Author)

  12. Materials for low-temperature fuel cells

    CERN Document Server

    Ladewig, Bradley; Yan, Yushan; Lu, Max

    2014-01-01

    There are a large number of books available on fuel cells; however, the majority are on specific types of fuel cells such as solid oxide fuel cells, proton exchange membrane fuel cells, or on specific technical aspects of fuel cells, e.g., the system or stack engineering. Thus, there is a need for a book focused on materials requirements in fuel cells. Key Materials in Low-Temperature Fuel Cells is a concise source of the most important and key materials and catalysts in low-temperature fuel cells. A related book will cover key materials in high-temperature fuel cells. The two books form part

  13. Materials for high-temperature fuel cells

    CERN Document Server

    Jiang, San Ping; Lu, Max

    2013-01-01

    There are a large number of books available on fuel cells; however, the majority are on specific types of fuel cells such as solid oxide fuel cells, proton exchange membrane fuel cells, or on specific technical aspects of fuel cells, e.g., the system or stack engineering. Thus, there is a need for a book focused on materials requirements in fuel cells. Key Materials in High-Temperature Fuel Cells is a concise source of the most important and key materials and catalysts in high-temperature fuel cells with emphasis on the most important solid oxide fuel cells. A related book will cover key mater

  14. Fuel performance, design and development

    International Nuclear Information System (INIS)

    Prasad, P.N.; Tripathi, Rahul Mani; Soni, Rakesh; Ravi, M.; Vijay Kumar, S.; Dwivedi, K.P.; Pandarinathan, P.R.; Neema, L.K.

    2006-01-01

    The normal fuel configurations for operating 220 MWe and 540 MWe PHWRs are natural uranium dioxide 19-element and 37- element fuel bundle types respectively. The fuel configuration for BWRs is 6 x 6 fuel. So far, about 330 thousand PHWR fuel bundles and 3500 number of BWR bundles have been irradiated in the 14 PHWRs and 2 BWRs. Improvements in fuel design, fabrication, quality control and operating practices are continuously carried out towards improving fuel utilization as well as reducing fuel failure rate. Efforts have been put to improve the fuel bundle utilization by increasing the fuel discharge burnup of the natural uranium bundles The overall fuel failure rate currently is less than 0.1 % . Presently the core discharge burnups in different reactors are around 7500 MWD/TeU. The paper gives the fuel performance experience over the years in the different power reactors and actions taken to improve fuel performance over the years. (author)

  15. Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

    Energy Technology Data Exchange (ETDEWEB)

    Yacout, A. M. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2016-09-16

    The US sodium cooled fast reactor (SFR) metallic fuel performance data that are of interest to advanced fast reactors applications, can be attributed mostly to the Integral Fast Reactor (IFR) program between 1984 and 1994. Metallic fuel data collected prior to the IFR program were associated with types of fuel that are not of interest to future advanced reactors deployment (e.g., previous U-Fissium alloy fuel). The IFR fuels data were collected from irradiation of U-Zr based fuel alloy, with and without Pu additions, and clad in different types of steels, including HT9, D9, and 316 stainless-steel. Different types of data were generated during the program, and were based on the requirements associated with the DOE Advanced Liquid Metal Cooled Reactor (ALMR) program.

  16. Handbook on process and chemistry on nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Atsuyuki (ed.) [Tokyo Univ., Tokyo (Japan); Asakura, Toshihide; Adachi, Takeo (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2001-12-01

    'Wet-type' nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of 'wet-type' reprocessing, because it contributes to establish and develop fuel reprocessing process and nuclear fuel cycle treating high burn-up UO{sub 2} fuel and spent MOX fuel, and to utilize 'wet-type' reprocessing technology much widely. This handbook summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing', from FY 1993 until FY 2000. (author)

  17. The battle of the fuel cell. De slag om de brandstofcel

    Energy Technology Data Exchange (ETDEWEB)

    Van Dijkum, P H [Nederlandse Organisatie voor Energie en Milieu BV (NOVEM), Sittard (Netherlands)

    1992-03-01

    An overview is given of several types of fuel cells and for each type the international state of the art in the development and technology. The fuel cells discussed are: the alkaline fuel cell (AFC), the proton exchange membrane fuel cell (PEMFC), the phosphoric acid fuel cell (PAFC), the external reforming molten carbonate fuel cell (ER-MCFC), the internal reforming molten carbonate fuel cell (IR-MCFC) and the solid oxide fuel cell (SOFC). 1 figs., 3 ills., 5 tabs., 7 refs.

  18. Laboratory measurements of trace gas emissions from biomass burning of fuel types from the southeastern and southwestern United States

    Science.gov (United States)

    Burling, I. R.; Yokelson, R. J.; Griffith, D. W. T.; Johnson, T. J.; Veres, P.; Roberts, J. M.; Warneke, C.; Urbanski, S. P.; Reardon, J.; Weise, D. R.; Hao, W. M.; de Gouw, J.

    2010-11-01

    Vegetation commonly managed by prescribed burning was collected from five southeastern and southwestern US military bases and burned under controlled conditions at the US Forest Service Fire Sciences Laboratory in Missoula, Montana. The smoke emissions were measured with a large suite of state-of-the-art instrumentation including an open-path Fourier transform infrared (OP-FTIR) spectrometer for measurement of gas-phase species. The OP-FTIR detected and quantified 19 gas-phase species in these fires: CO2, CO, CH4, C2H2, C2H4, C3H6, HCHO, HCOOH, CH3OH, CH3COOH, furan, H2O, NO, NO2, HONO, NH3, HCN, HCl, and SO2. Emission factors for these species are presented for each vegetation type burned. Gas-phase nitrous acid (HONO), an important OH precursor, was detected in the smoke from all fires. The HONO emission factors ranged from 0.15 to 0.60 g kg-1 and were higher for the southeastern fuels. The fire-integrated molar emission ratios of HONO (relative to NOx) ranged from approximately 0.03 to 0.20, with higher values also observed for the southeastern fuels. The majority of non-methane organic compound (NMOC) emissions detected by OP-FTIR were oxygenated volatile organic compounds (OVOCs) with the total identified OVOC emissions constituting 61 ± 12% of the total measured NMOC on a molar basis. These OVOC may undergo photolysis or further oxidation contributing to ozone formation. Elevated amounts of gas-phase HCl and SO2 were also detected during flaming combustion, with the amounts varying greatly depending on location and vegetation type. The fuels with the highest HCl emission factors were all located in the coastal regions, although HCl was also observed from fuels farther inland. Emission factors for HCl were generally higher for the southwestern fuels, particularly those found in the chaparral biome in the coastal regions of California.

  19. Composition heterogeneity analysis for DUPIC fuel(I) - Statistical analysis

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-08-01

    The fuel composition heterogeneity effect on reactor performance parameters was assessed by refueling simulations for three DUPIC fuel options of fuel composition heterogeneity control: the fissile content adjustment, the reactivity control by slightly enriched and depleted uranium, and the reactivity control by natural uranium. For each DUPIC fuel option, the simulations were performed using 30 heterogeneous fuel types which were determined by the agglomerative hierarchical clustering method. The heterogeneity effect was considered during the refueling simulation by randomly selecting fuel types for the refueling operation. The refueling simulations of the heterogeneous core have shown that the key performance parameters such as the maximum channel power (MCP), maximum bundle power (MBP), and channel power peaking factor (CPPF) are close to those of the core that has single fuel type. For the three DUPIC fuel options, the uncertainties of MCP, MBP, and CPPF due to the fuel composition heterogeneity are less than 0.6, 1.5 and 0.8%, respectively, including the uncertainty of the group-average fuel property. This study has shown that the three DUPIC fuel options reduces the composition heterogeneity effectively and the zone power control system has a sufficient margin to adjust the perturbations cased by the fuel composition heterogeneity. 15 refs., 28 figs.,10 tabs. (Author)

  20. Nuclear fuel element

    International Nuclear Information System (INIS)

    Mogard, J.H.

    1977-01-01

    A nuclear fuel element is disclosed for use in power producing nuclear reactors, comprising a plurality of axially aligned ceramic cylindrical fuel bodies of the sintered type, and a cladding tube of metal or metal alloys, wherein said cladding tube on its cylindrical inner surface is provided with a plurality of slightly protruding spacing elements distributed over said inner surface

  1. Nuclear fuel storage facility

    International Nuclear Information System (INIS)

    Matsumoto, Takashi; Isaka, Shinji.

    1987-01-01

    Purpose: To increase the spent fuel storage capacity and reduce the installation cost in a nuclear fuel storage facility. Constitution: Fuels handled in the nuclear fuel storage device of the present invention include the following four types: (1) fresh fuels, (2) 100 % reactor core charged fuels, (3) spent fuels just after taking out and (4) fuels after a certain period (for example one half-year) from taking out of the reactor. Reactivity is high for the fuels (1), and some of fuels (2), while low in the fuels (3) (4), Source intensity is strong for the fuels (3) and some of the fuels (2), while it is low for the fuels (1) and (4). Taking notice of the fact that the reactivity, radioactive source intensity and generated after heat are different in the respective fuels, the size of the pool and the storage capacity are increased by the divided storage control. While on the other hand, since the division is made in one identical pool, the control method becomes important, and the working range is restricted by means of a template, interlock, etc., the operation mode of the handling machine is divided into four, etc. for preventing errors. (Kamimura, M.)

  2. Degradation of solid oxide fuel cell metallic interconnects in fuels containing sulfur

    Energy Technology Data Exchange (ETDEWEB)

    Ziomek-Moroz, M.; Hawk, Jeffrey A.

    2005-01-01

    Hydrogen is the main fuel for all types of fuel cells except direct methanol fuel cells. Hydrogen can be generated from all manner of fossil fuels, including coal, natural gas, diesel, gasoline, other hydrocarbons, and oxygenates (e.g., methanol, ethanol, butanol, etc.). Impurities in the fuel can cause significant performance problems and sulfur, in particular, can decrease the cell performance of fuel cells, including solid oxide fuel cells (SOFC). In the SOFC, the high (800-1000°C) operating temperature yields advantages (e.g., internal fuel reforming) and disadvantages (e.g., material selection and degradation problems). Significant progress in reducing the operating temperature of the SOFC from ~1000 ºC to ~750 ºC may allow less expensive metallic materials to be used for interconnects and as balance of plant (BOP) materials. This paper provides insight on the material performance of nickel, ferritic steels, and nickel-based alloys in fuels containing sulfur, primarily in the form of H2S, and seeks to quantify the extent of possible degradation due to sulfur in the gas stream.

  3. Fuel cladding mechanical interaction during power ramps

    International Nuclear Information System (INIS)

    Guerin, Y.

    1985-01-01

    Mechanical interaction between fuel and cladding may occur as a consequence of two types of phenomenon: i) fuel swelling especially at levels of caesium accumulation, and ii) thermal differential expansion during power changes. Slow overpower ramps which may occur during incidental events are of course one of the circumstances responsible for this second type of fuel cladding mechanical interaction (FCMI). Experiments and analysis of this problem that have been done at C.E.A. allow to determine the main parameters which will fix the level of stress and the risk of damage induced by the fuel in the cladding during overpower transients

  4. Guide for subdivision of spent fuel pool. Project UNESA MAAP5-SFP; Guia para subdivision de la piscina de combustible gastado. Proyecto UNESA MAAP5-SFP

    Energy Technology Data Exchange (ETDEWEB)

    Martinez Barrios, M.; Garcia Gonzalez, M.; Perez Martin, F. J.

    2013-07-01

    The main goal of the UNESA MAAP5-SFP project is to analyze the capabilities of MAAP5 code and, particularly, the Spent Fuel Pool (SFP) module in order to tackle its modeling and facilitate the development of specific SFP models of Spanish NPPs. Within the project, Empresarios Agrupados (EEAA) is the responsible for the development of the Guide for the subdivision of the Spent Fuel Pool (SFP). This Guide includes a theoretical description of the model that is used by the code and a sequence of practical cases with the aim to evaluate the influence of specific parameters.

  5. Prototypical consolidation demonstration project - Final fuel recommendation report

    International Nuclear Information System (INIS)

    Piscitella, R.R.; Paskey, W.R.

    1987-01-01

    The Prototypical Consolidation Demonstration (PCD) Project will, in its final phase, conduct a demonstration of the equipment's ability to consolidate actual spent commercial fuel. Since budget and schedule limitations do not allow this demonstration to include all types of fuel assemblies, a selection process was utilized to identify the fuel types that would represent predominate fuel inventories and that would demonstrate the equipment's abilities. The Pressurized Water Reactor (PWR) fuel assemblies that were suggested for use in the PCD Project Hot Demonstration were Babcock and Wilcox (B and W) 15 x 15's, and Westinghouse (WE) 15 x 15's. The Boiling Water Reactor (BWR) fuel suggested was the General Electric (GE) 8 x 8

  6. The analysis of coolant-velocity distribution in plat-typed fuel element using CFD method for RSG-GAS research reactor

    International Nuclear Information System (INIS)

    Muhammad Subekti; Darwis Isnaini; Endiah Puji Hastuti

    2013-01-01

    The measurement experiment for coolant-velocity distribution in the subchannel of fuel element of RSG-GAS research reactor is difficult to be carried out due to too narrow channel and subchannel placed inside the fuel element. Hence, the calculation is required to predict the coolant-velocity distribution inside subchannel to confirm that the handle presence does not ruin the velocity distribution into every subchannel. This calculation utilizes CFD method, which respect to 3-dimension interior. Moreover, the calculation of coolant-velocity distribution inside subchannel was not ever carried out. The research object is to investigate the distribution of coolant-velocity in plat-typed fuel element using 3-dimension CFD method for RSG-GAS research reactor. This research is required as a part of the development of thermalhydraulic design of fuel element for innovative research reactor as well. The modeling uses ½ model in Gambit software and calculation uses turbulence equation in FLUENT 6.3 software. Calculation result of 3D coolant-velocity in subchannel using CFD method is lower about 4.06 % than 1D calculation result due to 1D calculation obeys handle availability. (author)

  7. BNFL's new spent fuel transport flask - Excellox 8

    International Nuclear Information System (INIS)

    McWilliam, D.S.

    2002-01-01

    Since British Nuclear Fuels plc (BNFL) was formed in 1971 its transport service has safely moved spent light water reactor fuel from many locations abroad to its fuel handling plants at Sellafield in the UK. To support this business a number of types of flasks have been designed and used. One of the types used has been the Excellox family of water-filled flasks. To support future business opportunities a new flask, designed to meet the requirements of the new IAEA transport regulations TS-R-1 (ST-1, Revised), has been developed. The flask will be a type B(U)F. This new flask design will maximise fuel carrying capacity to minimise transport costs. The design capacity of the new Excellox 8 flask is to be 12 pressurised water reactor or 32 boiling water reactor fuel assemblies. The objective of this BNFL project is to provide another economic spent nuclear fuel transport system, in support of BNFL transport business. (author)

  8. Fuel Cell Handbook, Fifth Edition

    Energy Technology Data Exchange (ETDEWEB)

    Energy and Environmental Solutions

    2000-10-31

    Progress continues in fuel cell technology since the previous edition of the Fuel Cell Handbook was published in November 1998. Uppermost, polymer electrolyte fuel cells, molten carbonate fuel cells, and solid oxide fuel cells have been demonstrated at commercial size in power plants. The previously demonstrated phosphoric acid fuel cells have entered the marketplace with more than 220 power plants delivered. Highlighting this commercial entry, the phosphoric acid power plant fleet has demonstrated 95+% availability and several units have passed 40,000 hours of operation. One unit has operated over 49,000 hours. Early expectations of very low emissions and relatively high efficiencies have been met in power plants with each type of fuel cell. Fuel flexibility has been demonstrated using natural gas, propane, landfill gas, anaerobic digester gas, military logistic fuels, and coal gas, greatly expanding market opportunities. Transportation markets worldwide have shown remarkable interest in fuel cells; nearly every major vehicle manufacturer in the U.S., Europe, and the Far East is supporting development. This Handbook provides a foundation in fuel cells for persons wanting a better understanding of the technology, its benefits, and the systems issues that influence its application. Trends in technology are discussed, including next-generation concepts that promise ultrahigh efficiency and low cost, while providing exceptionally clean power plant systems. Section 1 summarizes fuel cell progress since the last edition and includes existing power plant nameplate data. Section 2 addresses the thermodynamics of fuel cells to provide an understanding of fuel cell operation at two levels (basic and advanced). Sections 3 through 8 describe the six major fuel cell types and their performance based on cell operating conditions. Alkaline and intermediate solid state fuel cells were added to this edition of the Handbook. New information indicates that manufacturers have stayed

  9. Alternatives to traditional transportation fuels: An overview

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    This report presents the first compilation by the Energy Information Administration (EIA) of information on alternatives to gasoline and diesel fuel. The purpose of the report is: (1) to provide background information on alternative transportation fuels and replacement fuels compared with gasoline and diesel fuel, and (2) to furnish preliminary estimates of alternative transportation fuels and alternative fueled vehicles as required by the Energy Policy Act of 1992 (EPACT), Title V, Section 503, ``Replacement Fuel Demand Estimates and Supply Information.`` Specifically, Section 503 requires the EIA to report annually on: (1) the number and type of alternative fueled vehicles in existence the previous year and expected to be in use the following year, (2) the geographic distribution of these vehicles, (3) the amounts and types of replacement fuels consumed, and (4) the greenhouse gas emissions likely to result from replacement fuel use. Alternative fueled vehicles are defined in this report as motorized vehicles licensed for on-road use, which may consume alternative transportation fuels. (Alternative fueled vehicles may use either an alternative transportation fuel or a replacement fuel.) The intended audience for the first section of this report includes the Secretary of Energy, the Congress, Federal and State agencies, the automobile manufacturing industry, the transportation fuel manufacturing and distribution industries, and the general public. The second section is designed primarily for persons desiring a more technical explanation of and background for the issues surrounding alternative transportation fuels.

  10. Radiation resistance of pyrocarbon-boned fuel and absorbing elements for HTGR

    International Nuclear Information System (INIS)

    Gurin, V.A.; Konotop, Yu.F.; Odejchuk, N.P.; Shirochenkov, S.D.; Yakovlev, V.K.; Aksenov, N.A.; Kuprienko, V.A.; Lebedev, I.G.; Samsonov, B.V.

    1990-01-01

    In choosing the reactor type, problems of nuclear and radiation safety are outstanding. The analysis of the design and experiments show that HTGR type reactors helium cooled satisfy all the safety requirements. It has been planned in the Soviet Union to construct two HTGR plants, VGR-50 and VG-400. Later it was decided to construct an experimental plant with a low power high temperature reactor (VGM). Spherical uranium-graphite fuel elements with coated fuel particles are supposed to be used in HTGR core. A unique technology for producing spherical pyrocarbon-bound fuel and absorbing elements of monolithic type has been developed. Extended tests were done to to investigate fuel elements behaviour: radiation resistance of coated fuel particles with different types of fuel; influence of the coated fuel particles design on gaseous fission products release; influence of non-sphericity on coated fuel particle performance; dependence of gaseous fission products release from fuel elements on the thickness of fuel-free cans; confining role of pyrocarbon as a factor capable of diminishing the rate of fission products release; radiation resistance of spherical fuel elements during burnup; radiation resistance of spherical absorbing elements to fast neutron fluence and boron burnup

  11. Early detection of coolant boiling in research reactors with MTR-type fuel

    International Nuclear Information System (INIS)

    Kozma, R.; Turkcan, E.; Verhoef, J.P.

    1992-10-01

    A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs

  12. Fuel manufacturing and utilization

    International Nuclear Information System (INIS)

    2005-01-01

    The efficient utilisation of nuclear fuel requires manufacturing facilities capable of making advanced fuel types, with appropriate quality control. Once made, the use of such fuels requires a proper understanding of their behaviour in the reactor environment, so that safe operation for the design life can be achieved. The International Atomic Energy Agency supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle. It provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection. The IAEA supports the development fuel modelling expertise in Member States, covering both normal operation and postulated and severe accident conditions. It provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation. The IAEA supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, it provides information and support research into the basic properties of fuel materials, including UO 2 , MOX and zirconium alloys. It further offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology

  13. Alternative Fuel Guidelines for Alternative Transportation Systems.

    Science.gov (United States)

    2011-01-31

    The Volpe Center documented the increased use of alternative fuels on vehicles owned and operated by federal land management agencies. For each alternative fuel type, the Volpe Center documented the availability of vehicles, fueling mechanisms and pr...

  14. Thermodynamic analysis of biofuels as fuels for high temperature fuel cells

    Science.gov (United States)

    Milewski, Jarosław; Bujalski, Wojciech; Lewandowski, Janusz

    2013-02-01

    Based on mathematical modeling and numerical simulations, applicativity of various biofuels on high temperature fuel cell performance are presented. Governing equations of high temperature fuel cell modeling are given. Adequate simulators of both solid oxide fuel cell (SOFC) and molten carbonate fuel cell (MCFC) have been done and described. Performance of these fuel cells with different biofuels is shown. Some characteristics are given and described. Advantages and disadvantages of various biofuels from the system performance point of view are pointed out. An analysis of various biofuels as potential fuels for SOFC and MCFC is presented. The results are compared with both methane and hydrogen as the reference fuels. The biofuels are characterized by both lower efficiency and lower fuel utilization factors compared with methane. The presented results are based on a 0D mathematical model in the design point calculation. The governing equations of the model are also presented. Technical and financial analysis of high temperature fuel cells (SOFC and MCFC) are shown. High temperature fuel cells can be fed by biofuels like: biogas, bioethanol, and biomethanol. Operational costs and possible incomes of those installation types were estimated and analyzed. A comparison against classic power generation units is shown. A basic indicator net present value (NPV) for projects was estimated and commented.

  15. Thermodynamic analysis of biofuels as fuels for high temperature fuel cells

    Directory of Open Access Journals (Sweden)

    Milewski Jarosław

    2013-02-01

    Full Text Available Based on mathematical modeling and numerical simulations, applicativity of various biofuels on high temperature fuel cell performance are presented. Governing equations of high temperature fuel cell modeling are given. Adequate simulators of both solid oxide fuel cell (SOFC and molten carbonate fuel cell (MCFC have been done and described. Performance of these fuel cells with different biofuels is shown. Some characteristics are given and described. Advantages and disadvantages of various biofuels from the system performance point of view are pointed out. An analysis of various biofuels as potential fuels for SOFC and MCFC is presented. The results are compared with both methane and hydrogen as the reference fuels. The biofuels are characterized by both lower efficiency and lower fuel utilization factors compared with methane. The presented results are based on a 0D mathematical model in the design point calculation. The governing equations of the model are also presented. Technical and financial analysis of high temperature fuel cells (SOFC and MCFC are shown. High temperature fuel cells can be fed by biofuels like: biogas, bioethanol, and biomethanol. Operational costs and possible incomes of those installation types were estimated and analyzed. A comparison against classic power generation units is shown. A basic indicator net present value (NPV for projects was estimated and commented.

  16. Fast breeder fuel element development

    International Nuclear Information System (INIS)

    Marth, W.; Muehling, G.

    1983-08-01

    This report is a compilation of the papers which have been presented during a seminar ''Fast Breeder Fuel Element Development'' held on November 15/16, 1982 at KfK. The papers give a survey of the status, of the obtained results and of the necessary work, which still has to be done in the frame of various development programmes for fast breeder fuel elements. In detail the following items were covered by the presentations: - the requirements and boundary conditions for the design of fuel pins and elements both for the reference concept of the SNR 300 core and for the large, commercial breeder type of the future (presentation 1,2 and 6); - the fabrication, properties and characterization of various mixed oxide fuel types (presentations 3,4 and 5); - the operational fuel pin behaviour, limits of different design concepts and possible mechanism for fuel pin failures (presentations (7 and 8); - the situation of cladding- and wrapper materials development especially with respect to the high burn-up values of commercial reactors (presentations 9 and 10); - the results of the irradiation experiments performed under steady-state and non-stationary operational conditions and with failed fuel pins (presentations 11, 12, 13 and 14). (orig./RW) [de

  17. Fuel cell catalyst degradation

    DEFF Research Database (Denmark)

    Arenz, Matthias; Zana, Alessandro

    2016-01-01

    Fuel cells are an important piece in our quest for a sustainable energy supply. Although there are several different types of fuel cells, the by far most popular is the proton exchange membrane fuel cell (PEMFC). Among its many favorable properties are a short start up time and a high power density...... increasing focus. Activity of the catalyst is important, but stability is essential. In the presented perspective paper, we review recent efforts to investigate fuel cell catalysts ex-situ in electrochemical half-cell measurements. Due to the amount of different studies, this review has no intention to give...

  18. Fuel quality issues in stationary fuel cell systems.

    Energy Technology Data Exchange (ETDEWEB)

    Papadias, D.; Ahmed, S.; Kumar, R. (Chemical Sciences and Engineering Division)

    2012-02-07

    Fuel cell systems are being deployed in stationary applications for the generation of electricity, heat, and hydrogen. These systems use a variety of fuel cell types, ranging from the low temperature polymer electrolyte fuel cell (PEFC) to the high temperature solid oxide fuel cell (SOFC). Depending on the application and location, these systems are being designed to operate on reformate or syngas produced from various fuels that include natural gas, biogas, coal gas, etc. All of these fuels contain species that can potentially damage the fuel cell anode or other unit operations and processes that precede the fuel cell stack. These detrimental effects include loss in performance or durability, and attenuating these effects requires additional components to reduce the impurity concentrations to tolerable levels, if not eliminate the impurity entirely. These impurity management components increase the complexity of the fuel cell system, and they add to the system's capital and operating costs (such as regeneration, replacement and disposal of spent material and maintenance). This project reviewed the public domain information available on the impurities encountered in stationary fuel cell systems, and the effects of the impurities on the fuel cells. A database has been set up that classifies the impurities, especially in renewable fuels, such as landfill gas and anaerobic digester gas. It documents the known deleterious effects on fuel cells, and the maximum allowable concentrations of select impurities suggested by manufacturers and researchers. The literature review helped to identify the impurity removal strategies that are available, and their effectiveness, capacity, and cost. A generic model of a stationary fuel-cell based power plant operating on digester and landfill gas has been developed; it includes a gas processing unit, followed by a fuel cell system. The model includes the key impurity removal steps to enable predictions of impurity breakthrough

  19. Electrochemical power sources batteries, fuel cells, and supercapacitors

    CERN Document Server

    Bagotsky, Vladimir S; Volfkovich, Yurij M

    2015-01-01

    Electrochemical Power Sources (EPS) provides in a concise way theoperational features, major types, and applications of batteries,fuel cells, and supercapacitors Details the design, operational features, andapplications of batteries, fuel cells, and supercapacitors Covers improvements of existing EPSs and thedevelopment of new kinds of EPS as the results of intense R&Dwork Provides outlook for future trends in fuel cells andbatteries Covers the most typical battery types, fuel cells andsupercapacitors; such as zinc-carbon batteries, alkaline manganesedioxide batteries, mercury-zinc cells, lead

  20. Recommendations guide for the correct prescription's tests of diagnosis by image; Guia de recomendaciones para la correcta solicitud de pruebas de diagnostico por imagen

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    The purpose of this guide is to protect the health of the population against the dangers of ionizing radiation. The unjustified studies involving exposure to radiation unnecessarily increases the cancer risk in the exposed population. Medical irradiations are the most important contribution to human exposure to ionizing radiation from artificial character. Since statistics indicate a growing trend in the number of practices as well as the number of facilities, the quality of those practices, its justification and its optimization are today a relevant topic to scientific societies and regulatory agencies. Since the adoption of the directive 97/43/EURATOM, and the realization of the Conference of Malaga most European countries have implemented action plans for radiation protection of patients, including the search for consensus on the optimization dose criteria and justification for the indication of the examinations. An example is the PR/118 guide for diagnosis studies requests. In this framework, two Conferences on Radiation Protection of Patient were held in our country in which 'Working Groups' were organized on the following topics: Radiodiagnosis, Radiotherapy, Nuclear Medicine and Radioprotection of the Pregnant Woman. Systematic actions that received strong institutional support of the Argentine Society of Radiology began and ended with the elaboration of a 'Program for Radiological Protection Patient' which is being implemented. [Spanish] El objetivo de esta Guia es la proteccion de la salud de la poblacion frente a los riesgos derivados de las radiaciones ionizantes. La realizacion injustificada de estudios que impliquen exposicion a la radiacion aumenta innecesariamente el riesgo de cancer en la poblacion expuesta. Las irradiaciones medicas son la contribucion mas importante a la exposicion humana a las radiaciones ionizantes de caracter artificial. Ya que las estadisticas indican una tendencia creciente en el numero de practicas asi como la cantidad de

  1. Preparation of U3O8 powder for MTR type fuel from ammonium uranyl carbonate

    International Nuclear Information System (INIS)

    Marcondes, G.H.; Riella, H.G.

    1990-08-01

    In this paper it is described the research done at IPEN-CNEN/SP on the preparation of U 3 O 8 powder from calcination of the AUC, with appropriate characteristics to be used as dispersoid for MTR type fuel. The calcination in air of the AUC leads a U 3 O 8 powder that is further processed to obtain a powder with density and particle size as especifications. The important process parameters are here discussed with the variation AUC calcination temperature and sintering time of the U 3 O 8 powder. (author) [pt

  2. Advanced fuel technology and performance

    International Nuclear Information System (INIS)

    1985-10-01

    The purpose of the Advisory Group Meeting on Advanced Fuel Technology and Performance was to review the experience of advanced fuel fabrication technology, its performance, peculiarities of the back-end of the nuclear fuel cycle with regard to all types of reactors and to outline the future trends. As a result of the meeting recommendations were made for the future conduct of work on advanced fuel technology and performance. A separate abstract was prepared for each of the 20 papers in this issue

  3. IRF3 and type I interferons fuel a fatal response to myocardial infarction.

    Science.gov (United States)

    King, Kevin R; Aguirre, Aaron D; Ye, Yu-Xiang; Sun, Yuan; Roh, Jason D; Ng, Richard P; Kohler, Rainer H; Arlauckas, Sean P; Iwamoto, Yoshiko; Savol, Andrej; Sadreyev, Ruslan I; Kelly, Mark; Fitzgibbons, Timothy P; Fitzgerald, Katherine A; Mitchison, Timothy; Libby, Peter; Nahrendorf, Matthias; Weissleder, Ralph

    2017-12-01

    Interferon regulatory factor 3 (IRF3) and type I interferons (IFNs) protect against infections and cancer, but excessive IRF3 activation and type I IFN production cause autoinflammatory conditions such as Aicardi-Goutières syndrome and STING-associated vasculopathy of infancy (SAVI). Myocardial infarction (MI) elicits inflammation, but the dominant molecular drivers of MI-associated inflammation remain unclear. Here we show that ischemic cell death and uptake of cell debris by macrophages in the heart fuel a fatal response to MI by activating IRF3 and type I IFN production. In mice, single-cell RNA-seq analysis of 4,215 leukocytes isolated from infarcted and non-infarcted hearts showed that MI provokes activation of an IRF3-interferon axis in a distinct population of interferon-inducible cells (IFNICs) that were classified as cardiac macrophages. Mice genetically deficient in cyclic GMP-AMP synthase (cGAS), its adaptor STING, IRF3, or the type I IFN receptor IFNAR exhibited impaired interferon-stimulated gene (ISG) expression and, in the case of mice deficient in IRF3 or IFNAR, improved survival after MI as compared to controls. Interruption of IRF3-dependent signaling resulted in decreased cardiac expression of inflammatory cytokines and chemokines and decreased inflammatory cell infiltration of the heart, as well as in attenuated ventricular dilation and improved cardiac function. Similarly, treatment of mice with an IFNAR-neutralizing antibody after MI ablated the interferon response and improved left ventricular dysfunction and survival. These results identify IRF3 and the type I IFN response as a potential therapeutic target for post-MI cardioprotection.

  4. Shippingport LWBR (Th/U Oxide) Fuel Characteristics for Disposal Criticality Analysis

    International Nuclear Information System (INIS)

    Taylor, L. L.; Loo, H. H.

    1999-01-01

    Department of Energy (DOE)-owned spent nuclear fuels encompass many fuel types. In an effort to facilitate criticality analysis for these various fuel types, they were categorized into eight characteristic fuel groups with emphasis on fuel matrix composition. Out of each fuel group, a representative fuel type was chosen for analysis as a bounding case within that fuel group. Generally, burnup data, fissile enrichments, and total fuel and fissile mass govern the selection of the representative or candidate fuel within that group. The Shippingport Light Water Breeder Reactor (LWBR) fuels incorporate more of the conventional materials (zirconium cladding/heavy metal oxides) and fabrication details (rods and spacers) that make them comparable to a typical commercial fuel assembly. The LWBR seed/blanket configuration tested a light-water breeder concept with Th-232/U-233 binary fuel matrix. Reactor design used several assembly configurations at different locations within the same core . The seed assemblies contain the greatest fissile mass per (displaced) unit volume, but the blanket assemblies actually contain more fissile mass in a larger volume; the atom-densities are comparable

  5. Suporte informacional como elemento para orientação de pais de pré-termo: um guia para o serviço de acompanhamento do desenvolvimento no primeiro ano de vida Information support as an element for advising parents of preterm infants: a guide for development follow-up services during the first year of life

    Directory of Open Access Journals (Sweden)

    CMS Martinez

    2007-02-01

    Full Text Available INTRODUÇÃO: A literatura aponta o alto risco de distúrbios biológicos e psicossociais aos quais o bebê pré-termo está susceptível, necessitando de atenção diferenciada daquela dada ao bebê a termo. OBJETIVO: Elaborar um guia de orientação e acompanhamento do desenvolvimento no primeiro ano de vida para pais que freqüentam um serviço de acompanhamento do desenvolvimento de bebês. MÉTODO: Os critérios para inclusão dos participantes foram: ser pais de bebê pré-termo de zero a seis meses de idade corrigida que não apresentavam doença neurológica e não necessitaram de intervenção fisioterápica. O presente estudo, de caráter qualitativo, empregou, na etapa de coleta de dados, a observação direta e a entrevista, efetivadas com as mães e profissionais deste serviço de acompanhamento do desenvolvimento de bebês. A elaboração do Guia baseou-se nas análises das observações dos atendimentos no serviço, na freqüência das respostas das entrevistas com profissionais e mães sobre necessidades e dúvidas no manuseio e cuidado com o bebê e a associação com a literatura. RESULTADOS: Obteve-se um guia contendo informações sobre o desenvolvimento do bebê, distribuído nos quatro primeiros trimestres de vida, e o alerta sobre a importância de calcular a idade corrigida para o acompanhamento adequado dos marcos do desenvolvimento. CONCLUSÃO: A utilização deste guia pode ser feita por diferentes profissionais da saúde e por aqueles que desenvolvem atividades educativas para pais.INTRODUCTION: The literature shows that preterm infants are at high risk of biological and psychological disorders and consequently require a higher level of care than is provided for full-term infants. OBJECTIVE: To draw up a guide advising on development follow-up during the first year of life, for parents who attend a child development follow-up service. METHOD: The inclusion criteria were that participants should be parents of preterm

  6. Strategies for fuel cell product development. Developing fuel cell products in the technology supply chain

    International Nuclear Information System (INIS)

    Hellman, H.L.

    2004-01-01

    Due to the high cost of research and development and the broad spectrum of knowledge and competences required to develop fuel cell products, many product-developing firms outsource fuel cell technology, either partly or completely. This article addresses the inter-firm process of fuel cell product development from an Industrial Design Engineering perspective. The fuel cell product development can currently be characterised by a high degree of economic and technical uncertainty. Regarding the technology uncertainty: product-developing firms are more often then not unfamiliar with fuel cell technology technology. Yet there is a high interface complexity between the technology supplied and the product in which it is to be incorporated. In this paper the information exchange in three current fuel cell product development projects is analysed to determine the information required by a product designer to develop a fuel cell product. Technology transfer literature suggests that transfer effectiveness is greatest when the type of technology (technology uncertainty) and the type of relationship between the technology supplier and the recipient are carefully matched. In this line of thinking this paper proposes that the information required by a designer, determined by the design strategy and product/system volume, should be met by an appropriate level of communication interactivity with a technology specialist. (author)

  7. Study of the Effect of Hydrocarbon Type Biodegradation on Fuel Specification Properties

    Science.gov (United States)

    2014-06-01

    diesel fuel (F10428) before and after 1 month exposure to Pseudomonas or a control. Figure 12. QCM profiles at 140°C of mass accumulation (solid...DLA-13) Figure 32. Calibration curve for analysis of BHT in jet fuel. Figure 33. Growth of yeast in 20 mg/L concentrations of A and B. Figure...bladder materials. Some costly problems associated with microbial growth include tank corrosion, fuel pump failures, filter plugging, injector

  8. World wide IFC phosphoric acid fuel cell implementation

    Energy Technology Data Exchange (ETDEWEB)

    King, J.M. Jr

    1996-04-01

    International Fuel Cells, a subsidary of United technologies Corporation, is engaged in research and development of all types of fuel cell technologies and currently manufactures alkaline fuel cell power plants for the U.S. manned space flight program and natural gas fueled stationary power plants using phosphoric acid fuel cells. This paper describes the phosphoric acid fuel cell power plants.

  9. Fuel saving type power plant for automobiles

    Energy Technology Data Exchange (ETDEWEB)

    Endo, N; Katsumoto, T; Shimizu, T; Hiramatsu, T; Fujita, Y

    1982-10-01

    Mitsubishi Motors Corporation has developed a modulated displacement engine named ''Orion MD'' and an electronically controlled damper clutch automatic transmission named ''ELC Automatic'' and has installed them on the new ''Mirage'' series and ''Cordia'' series, respectively, which were put on sale in February, 1982. They improve fuel economy to a great extent especially at low vehicle speed, and provide good driveability and high reliability. An outline of the ''Orion MD'' and ''ELC Automatic'' is presented.

  10. Fuel rods

    International Nuclear Information System (INIS)

    Adachi, Hajime; Ueda, Makoto

    1985-01-01

    Purpose: To provide a structure capable of measuring, in a non-destructive manner, the releasing amount of nuclear gaseous fission products from spent fuels easily and at a high accuracy. Constitution: In order to confirm the integrity and the design feasibility of a nuclear fuel rod, it is important to accurately determine the amount of gaseous nuclear fission products released from nuclear pellets. In a structure where a plurality of fuel pellets are charged in a fuel cladding tube and retained by an inconel spring, a hollow and no-sealed type spacer tube made of zirconium or the alloy thereof, for example, not containing iron, cobalt, nickel or manganese is formed between the spring and the upper end plug. In the fuel rod of such a structure, by disposing a gamma ray collimator and a gamma ray detector on the extension of the spacer pipe, the gamma rays from the gaseous nuclear fission products accumulated in the spacer pipe can be detected while avoiding the interference with the induction radioactivity from inconel. (Kamimura, M.)

  11. Main results and status of the development of LEU fuel for Russian research reactors

    International Nuclear Information System (INIS)

    Vatulin, A.; Morozov, A.; Suprun, V.; Dobrikova, I.

    2005-01-01

    VNIINM develops low enrichment uranium (LEU) fuel on base U-Mo alloys and a novel design of pin-type fuel elements. The development is carried out both for existing reactors, and for new advanced designs of reactors. The work is carried on the following main directions: - irradiate LEU U-Mo dispersion fuel (the uranium density up to 6,0 g/cm 3 ) in two Russian research reactors: MIR (RIAR, Dimitrovgrad) as pin type fuel mini-elements and in WWR-M (PINP, Gatchina) within full-scaled fuel assembly (FA) with pin type fuel elements; - finalize development of design and fabrication process of IRT type FA with pin type fuel elements; - develop methods of reducing of U-Mo fuel --Al matrix interaction under irradiation; - develop fabricating methods of fuel elements on base of monolithic U-Mo fuel. The paper generally reviews the results of calculation, design and technology investigations accomplished by now. (author)

  12. Licensing of new fuel for Slovak NPPs

    International Nuclear Information System (INIS)

    Melicharek, M.; Mikulas, B.

    2011-01-01

    With the hope of achieving a six year fuel cycle and to allow for further power uprates a new type of fuel was licensed for NPP Mochovce Unit 1 (EMO1). Slovenske elektrarne, a.s. introduced the idea for the new fuel type to the regulatory authority in the middle of 2010. That's when the first draft of EMO safety report was made. This was the actual start of the licensing process. In case of new fuel regulator has several requirements like extended start up and power test; introducing a fuel behavior monitoring plan; using two independent methods to determine coolant flow though the active zone. In this paper the following issues have been discussed: 1) Engineering and safety reserve factors; 2)Fuel limits; 3) Uncertainties decrease in measurement of coolant flow through the reactor; 4) Determination of heating on fuel assembly; 5) Storage pool issues; 6) Emergency planning; 7) Monitoring of several fuel parameters (K qmax , K rmax , K v .K k and bypass)monitored during campaign

  13. Fabrication of AA6061-T6 Plate Type Fuel Assembly Using Electron Beam Welding Process

    International Nuclear Information System (INIS)

    Kim, Soosung; Seo, Kyoungseok; Lee, Donbae; Park, Jongman; Lee, Yoonsang; Lee, Chongtak

    2014-01-01

    AA6061-T6 aluminum alloy is easily welded by conventional GTAW (Gas Tungsten Arc Welding), LBW (Laser Beam Welding) and EBW. However, certain characteristics, such as solidification cracking, porosity, HAZ (Heat-affected Zone) degradation must be considered during welding. Because of high energy density and low heat input, especially LBW and EBW processes possess the advantage of minimizing the fusing zone and HAZ and producing deeper penetration than arc welding processes. In present study, to apply for the nuclear fuel plate fabrication and assembly, a fundamental EBW experiment using AA6061-T6 aluminum alloy specimens was conducted. Furthermore, to establish the welding process, and satisfy the requirements of the weld quality, EBW apparatus using an electron welding gun and vacuum chamber was developed, and preliminary investigations for optimizing the welding parameters of the specimens using AA6061-T6 aluminum plates were also performed. The EB weld quality of AA6061-T6 aluminum alloy for the fuel plate assembly has been also studied by the shrinkage measurement and weld inspection using computed tomography. This study was carried out to determine the suitable welding parameters and to evaluate tensile strength of AA6061-T6 aluminum alloy. In the present experiment, satisfactory electron beam welding process of the full-sized sample was being developed. Based on this fundamental study, fabrication of the plate-type fuel assembly will be provided for the future Ki-Jang research reactor project

  14. Study on a New Type of Electric-controlled Engine Fuel Consumption Meter Based on Volume Method

    Directory of Open Access Journals (Sweden)

    Qing-Yong Zhang

    2014-04-01

    Full Text Available At present study on the testing methods and instruments for vehicles’ fuel consumption is still not perfect. It still can’t provide a rapid and accurate measuring method and instrument. A new type of fuel consumption meter structure is developed which used two small containers to relay to supply the engine and realizes oil consumption measuring by detecting the real- time liquid level in the containers. Photoelectric sensors and a chip microcomputer are used to realize transient detection. Its structure and principle are analyzed. The system of its hardware and software of the electric-controlling system are designed. Some key components are selected and the process of exhausting, starting and measuring are designed. Precision test of the system is performed, and the result shows the accuracy of the meter in the range of 800 ml is 0.1 %, which meets the requirements and the feasibility of the structure is verified. Finally the main influencing factors are analyzed.

  15. INPR ACPR utilization in fuel behaviour studies under accidental condition

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Popov, Mircea

    1990-01-01

    This paper is dedicated to the experimental program, investigating CANDU type fuel behaviour in transient condition, as well as the facilities supporting this program. The tests Reactivity Initiated Accident type. The experiments were performed within TRIGA ACPR facility, installed at INSTITUTE for NUCLEAR POWER REACTORS, Pitesti, ROMANIA. Studies of the safety issues took a great international developement during last years. In USA, Japan, owners of the similar reactors, and USSR there are a big commitment to such programs, intended to establish the nuclear fuel behaviour under RIA-conditions. In our country, too, there are programs aiming a complete testing of the CANDU type fuels. As it is known, RIA is not a CANDU specific accident, but the fuel behaviour in such conditions can give useful informations on the fuel cladding failure threshold and about reflooding post LOCA heat transfer condition. Based on some papers and specific requirements it was initiated and developed a safety research program on CANDU type fuel using the ACPR. The paper describes the reactor,test capsule, instrumentation, fuel samples, tests, post irradiation results. (orig.)

  16. Systematic study on Thorium fuel

    International Nuclear Information System (INIS)

    Shibata, Toshikazu; Kimura, Itsuro; Iwata, Shiro; Furuya, Hirotaka; Suzuki, Susumu.

    1988-01-01

    Introduced is the activities of the Joint Research Project Team on Thorium Fuel organized by mainly university researchers in Japan and supported by the Ministry of Education, Science and Culture for seven years since 1980. Four major groups were organized; (1) nuclear data, reactor physics and design, (2) nuclear fuel, (3) down stream and (4) biological effects of thorium. The first group covered measurements and analysis on nuclear data of thorium related nuclides, experiment and analysis on nuclear characteristics of thorium containing cores, basic engineering on a thorium molten salt reactor, and designs of several types of reactors. Fabrication and irradiation tests of thorium oxide fuel, and basic studies on new type thorium fuels (e.g. carbide and nitride) were studied by the second group. The third group covered the use of solutions in reprocessing of spent fuel, behavior of fission products, immobilization of high level radioactive waste, and continuous reprocessing for a molten salt reactor. The fourth group performed the trace study for patients who had been intravascularly injected with thorotrast for diagnosis of war injuries during the Second World War. (author)

  17. Fuel management of HTR-10

    International Nuclear Information System (INIS)

    Wu Zongxin; Jing Xingqing

    2001-01-01

    The 10 MW high temperature cooled reactor (HTR-10) built in Tsinghua University is a pebble bed type of HTGR. The continuous recharge and multiple-pass of spherical fuel elements are used for fuel management. The initiative stage of core is composed of the mix of spherical fuel elements and graphite elements. The equilibrium stage of core is composed of identical spherical fuel elements. The fuel management during the transition from the initiative stage to the equilibrium stage is a key issue for HTR-10 physical design. A fuel management strategy is proposed based on self-adjustment of core reactivity. The neutron physical code is used to simulate the process of fuel management. The results show that the graphite elements, the recharging fuel elements below the burn-up allowance, and the discharging fuel elements over the burn-up allowance could be identified by burn-up measurement. The maximum of burn-up fuel elements could be controlled below the burn-up limit

  18. HEU and Leu FueL Shielding Comparative Study Applied for Spent Fuel Transport

    International Nuclear Information System (INIS)

    Margeanu, C.A.; Margeanu, S.; Barbos, D.

    2009-01-01

    INR Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies. The dual core concept involves the operation of a 14 MW TRIGA steady-state, high flux research and material testing reactor at one end of a large pool, and the independent operation of an annular-core pulsing reactor (TRIGA-ACPR) at the other end of the pool. The steady-state reactor is mostly used for long term testing of power reactor fuel components (pellets, pins, subassemblies and fuel assemblies) followed by post-irradiation examination. Following the general trend to replace the He fuel type (High Enriched Uranium) by Leu fuel type (Low Enriched Uranium), in the light of international agreements between IAEA and the states using He fuel in their nuclear reactors, Inr Past's have been accomplished the TRIGA research reactor core full conversion on May 2006. The He fuel repatriation in US in the frame of Foreign Research Reactor Spent Nuclear Fuel Return Programme effectively started in 1999, the final stage being achieved in summer of 2008. Taking into account for the possible impact on the human and environment, in all activities associated to nuclear fuel cycle, the spent fuel or radioactive waste characteristics must be well known. Shielding calculations basic tasks consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper is a comparative study of Leu and He fuel utilization effects for the shielding analysis during spent fuel transport. A comparison against the measured data for He spent fuel, available from the last stage of the spent fuel repatriation, is presented. All the geometrical and material data related on the spent fuel shipping cask were considered according to the Nac-Lt Cask approved model. The shielding analysis estimates radiation doses to shipping cask wall surface

  19. Considerations for handling failed fuel at the Barnwell Nuclear Fuel Plant

    International Nuclear Information System (INIS)

    Anderson, R.T.; Cholister, R.J.

    1982-05-01

    The impact of failed fuel receipt on reprocessing operations is qualitatively described. It appears that extended storage of fuel, particularly with advanced storage techniques, will increase the quantity of failed fuel, the nature and possibly the configuration of the fuel. The receipt of failed fuel at the BNFP increases handling problems, waste volumes, and operator exposure. If it is necessary to impose special operating precautions to minimize this impact, a loss in plant throughput will result. Hence, ideally, the reprocessing plant operator would take every reasonable precaution so that no failed fuel is received. An alternative policy would be to require that failed fuel be placed in a sealed canister. In the latter case the canister must be compatible with the shipping cask and suitable for in-plant storage. A required inspection of bare fuel would be made at the reactor prior to shipping off-site. This would verify fuel integrity. These requirements are obviously idealistic. Due to the current uncertain status of reprocessing and the need to keep reactors operating, business or governmental policy may be enacted resulting in the receipt of a negotiated quantity of non-standard fuel (including failed fuel). In this situation, BNFP fuel receiving policy based soley on fuel cladding integrity would be difficult to enforce. There are certain areas where process incompatibility does exist and where a compromise would be virtually impossible, e.g., canned fuel for which material or dimensional conflicts exist. This fuel would have to be refused or the fuel would require recanning prior to shipment. In other cases, knowledge of the type and nature of the failure may be acceptable to the operator. A physical inspection of the fuel either before shipment or after the cask unloading operation would be warranted. In this manner, concerns with pool contamination can be identified and the assembly canned if deemed necessary

  20. Development and testing of a hybrid system with a sub-kW open-cathode type PEM (proton exchange membrane) fuel cell stack

    International Nuclear Information System (INIS)

    Huang, Zhen-Ming; Su, Ay; Liu, Ying-Chieh

    2014-01-01

    In this study, the performance of a polymer electrolyte membrane fuel cell stack has been evaluated for a hybrid power system test platform. To simulate vehicle acceleration, the stack was operated under dynamic-loading, and to demonstrate the exchange of power flow between two power sources the hybrid power system was tested under three different modes. A unit cell was fabricated for high stack performance and the stack was constructed with 18 open-cathode type fuel cells. Air which acts as a coolant as well as an oxidant for electrochemical reactions is provided by a pair of fans. The capabilities of the stack for hybrid power system test platform were validated by successful dynamic-loading tests. The performance of the stack for various air fan voltage was evaluated and an optimal value was concluded. The conditions like inlet temperature of H 2 and the stack current were established for maximum power. It was also found that humidification of hydrogen at anode inlet degrades the stack performance and stability due to flooding. Evidence shows that for the higher overall performance, the fuel cell acts continuously on constant current output. The study contributes to the design of mobility hybrid system to get better performance and reliability. - Highlights: • An open-cathode type PEMFC (polymer electrolyte membrane fuel cell) stack (rated output 300 W) was fabricated. • The open-cathode configuration simplifies the design of a stack system. • Assess the feasibility of combining a fuel cell stack in a hybrid system. • The study contributes to the design of mobility hybrid system to get better performance and reliability

  1. The birth of the fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Prohaska, Don

    2001-12-01

    Everyone knows that Thomas Alva Edison invented the light bulb, Alexander Graham Bell the telephone and that the Otto and Diesel engines were invented by two Germans bearing those names. But who invented the fuel cell? Fuel cells generate electricity with virtually zero pollution by combining gaseous fuels and air. There are different types generally described as high temperature or low temperature fuel cells. Here, Don Prohaska delves into a recently published book: The Birth of the Fuel Cell, by a descendant of one of the fathers of the fuel cell, and sheds new light on the early days of this technology. (Author)

  2. Reactor fuel charging equipment

    International Nuclear Information System (INIS)

    Wade, Elman.

    1977-01-01

    In many types of reactor fuel charging equipment, tongs or a grab, attached to a trolley, housed in a guide duct, can be used for withdrawing from the core a selected spent fuel assembly or to place a new fuel assembly in the core. In these facilities, the trolley may have wheels that roll on rails in the guide duct. This ensures the correct alignment of the grab, the trolley and fuel assembly when this fuel assembly is being moved. By raising or lowering such a fuel assembly, the trolley can be immerged in the coolant bath of the reactor, whereas at other times it can be at a certain level above the upper surface of the coolant bath. The main object of the invention is to create a fuel handling apparatus for a sodium cooled reactor with bearings lubricated by the sodium coolant and in which the contamination of these bearings is prevented [fr

  3. Bio-based production of fuels and industrial chemicals by repurposing antibiotic-producing type I modular polyketide synthases: opportunities and challenges

    DEFF Research Database (Denmark)

    Yuzawa, Satoshi; Keasling, Jay D.; Katz, Leonard

    2017-01-01

    Complex polyketides comprise a large number of natural products that have broad application in medicine and agriculture. They are produced in bacteria and fungi from large enzyme complexes named type I modular polyketide synthases (PKSs) that are composed of multifunctional polypeptides containin...... have applications as fuels or industrial chemicals....

  4. Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

    International Nuclear Information System (INIS)

    Arkoma, Asko; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

    2015-01-01

    Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

  5. Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Arkoma, Asko, E-mail: asko.arkoma@vtt.fi; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

    2015-04-15

    Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

  6. TMI-2 spent fuel shipping

    International Nuclear Information System (INIS)

    Quinn, G.J.; Burton, H.M.

    1985-01-01

    TMI-2 failed fuel will be shipped to the Idaho National Engineering Laboratory for use in the DOE Core Examination Program. The fuel debris will be loaded into three types of canisters during defueling and dry loaded into a spent fuel shipping cask. The cask design accommodates seven canisters per cask and has two separate containment vessels with ''leaktight'' seals. Shipments are expectd to begin in early 1986

  7. The fuel of nuclear reactors

    International Nuclear Information System (INIS)

    1995-03-01

    This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

  8. Development of remote controlled type field-emission type scanning electron microscope

    Energy Technology Data Exchange (ETDEWEB)

    Yasuda, Ryo; Nishino, Yasuharu; Mita, Naoaki; Nakata, Masahito; Harada, Katsuya; Nozawa, Yukio; Amano, Hidetoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-10-01

    The extending burn-up of Light Water Reactor Fuels has been promoted to reduce costs of the power generation and amount of waste mass. Information about the fuel behavior under high burn-up operation is needed to assess safety of the high burn-up fuels. Microstructures formed in high burn-up fuel pellets and Zircaloy tubes influence on their integrity. The fundamental information about morphology, sizes, and element compositions in those microstructures is necessary to estimate the formation mechanism and change in the properties of the fuels. The Field Emission type Scanning Electron Microscope (FE-SEM), which is effective for observation of very small area, i.e., nano-size structures, has been hence installed at the Reactor Fuel Examination Facility (RFEF) in JAERI. FE-SEM is designed for the remote handling type to use high radioactive materials and has equipments to keep the safety for operators. The Energy Dispersive Spectrometer (EDS) with a radiation-shielding collimator has been also equipped on FE-SEM to determine element compositions in the structures. Characterization tests were carried out using Zircaloy cladding tubes with oxide films and hydrides of confirm machine performance. In the results of the tests, high-resolution images with a magnification of 30,000 were obtained. Those results show that the apparatus maintains the original high performance with standard type. (author)

  9. Chemical characterization of nuclear fuel materials

    International Nuclear Information System (INIS)

    Ramakumar, K.L.

    2011-01-01

    India is fabricating nuclear fuels for various types of reactors, for example, (U-Pu) MOX fuel of varying Pu content for boiling water reactors (BWRs), pressurized heavy water reactors (PHWRs), prototype fast breeder reactors (PFBRs), (U-Pu) carbide fuel fast breeder test reactor (FBTR), and U-based fuels for research reactors. Nuclear fuel being the heart of the reactor, its chemical and physical characterisation is an important component of this design. Both the fuel materials and finished fuel products are to be characterised for this purpose. Quality control (both chemical and physical) provides a means to ensure that the quality of the fabricated fuel conforms to the specifications for the fuel laid down by the fuel designer. Chemical specifications are worked out for the major and minor constituents which affect the fuel properties and hence its performance under conditions prevailing in an operating reactor. Each fuel batch has to be subjected to comprehensive chemical quality control for trace constituents, stoichiometry and isotopic composition. A number of advanced process and quality control steps are required to ensure the quality of the fuels. Further more, in the case of Pu-based fuels, it is necessary to extract maximum quality data by employing different evaluation techniques which would result in minimum scrap/waste generation of valuable plutonium. The task of quality control during fabrication of nuclear fuels of various types is both challenging and difficult. The underlying philosophy is total quality control of the fuel by proper mix of process and quality control steps at various stages of fuel manufacture starting from the feed materials. It is also desirable to adapt more than one analytical technique to increase the confidence and reliability of the quality data generated. This is all the most required when certified reference materials are not available. In addition, the adaptation of non-destructive techniques in the chemical quality

  10. Improved techniques for appendage attachment to PHWR fuel elements

    International Nuclear Information System (INIS)

    Raj, R.N.J.; Laxminarayana, B.; Narayanan, P.S.A.; Gupta, U.C.; Varma, B.P.; Sinha, K.K.

    1995-01-01

    Nuclear Fuel Complex, India switched-over to split-wart type PHWR fuel bundles in mid-80s. Since then over 60,000 bundles of this type have been fabricated for Indian PHWRs. After considering various technical aspects, resistance welding was chosen for appendage attachment to the fuel elements. The paper describes experiences in scaling up of the technique to industrial production of PHWR fuel bundles, design and development of special-purpose equipment for this purpose, and the QA procedures employed for regular production. It also deals with appendage welding of 37 Element fuel bundles and improvements planned in the appendage welding process. (author)

  11. Fuel analyzer; Analisador de combustiveis

    Energy Technology Data Exchange (ETDEWEB)

    Cozzolino, Roberval [RS Motors, Indaiatuba, SP (Brazil)

    2008-07-01

    The current technology 'COMBUSTIMETRO' aims to examine the fuel through performance of the engine, as the role of the fuel is to produce energy for the combustion engine in the form of which is directly proportional to the quality and type of fuel. The 'COMBUSTIMETRO' has an engine that always keeps the same entry of air, fuel and fixed point of ignition. His operation is monitored by sensors (Sonda Lambda, RPM and Gases Analyzer) connected to a processor that performs calculations and records the information, generate reports and graphs. (author)

  12. Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Robinson; G. S. Chang; D. D. Keiser, Jr.; D. M. Wachs; D. L. Porter

    2009-08-01

    A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

  13. Interim spent-fuel storage options at commercial nuclear power plants

    International Nuclear Information System (INIS)

    Thakkar, A.R.; Hylko, J.M.

    1991-01-01

    Although spent fuel can be stored safely in waterfilled pools at reactor sites, some utilities may not possess sufficient space for life-of-plant storage capability. In-pool storage capability may be increased by reracking assemblies, rod consolidation, double tiering spent-fuel racks, and by shipping spent fuel to other utility-owned facilities. Long-term on-site storage capability for spent fuel may be provided by installing (dry-type) metal casks, storage and transportation casks, concrete casks, horizontal concrete modules, modular concrete vaults, or by constructing additional (pool-type) storage installations. Experience to date has provided valuable information regarding dry-type or pool-type installations, cask handling and staffing requirements, security features, decommissioning activities, and radiological issues

  14. Structural analysis on the open basket type instrumented capsule for fuel irradiation tests in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Sik; Kang, Y. H.; Kim, B. G.; Cho, M. S.; Sohn, J. M.; Choo, K. N.; Oh, J. M.; Shin, Y. T.; Park, S. J. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-05-01

    To develop the open basket type instrumented capsule to be used for the irradiation test of various nuclear fuels, it is necessary to ensure the compatibility of the capsule with HANARO and the structural integrity of the capsule. The dimensions of the open basket type instrumented capsule were determined in the basis of the pressure drop criteria in OR test hole of HANARO(mass flow rate <12.7kg/s, pressure drop {delta}P>200kPa). From the buckling stability analysis for this capsule, the critical buckling load P{sub cr} was 7.5kN. The vertical impact stress of the capsule under unit impact load was evaluated by the transient analysis, and the maximum vertical impact load calculated from the impact stress and the allowable stress was 60.5kN. Under the loading of the calculated Pcr, the maximum vertical impact stress was 20.4MPa. The structural integrity of the capsule under a horizontal impact loading was also examined. The mechanical stresses occurred by the pressure difference at the inner and outer surface of cladding and by the coolant pressure at the surface of cladding were 3.1MPa and 43.3MPa, respectively. These stress values were lower than the allowable stress in each case. Therefore, it was ensured that the instrumented capsule for the irradiation test of various nuclear fuels met the criteria on the structural integrity during installing and testing the capsule in HANARO. 8 refs., 61 figs., 3 tabs. (Author)

  15. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Sei; Ando, Ryohei; Mitsutake, Toru.

    1995-01-01

    The present invention concerns a fuel assembly suitable to a BWR-type reactor and improved especially with the nuclear characteristic, heat performance, hydraulic performance, dismantling or assembling performance and economical property. A part of poison rods are formed as a large-diameter/multi-region poison rods having a larger diameter than a fuel rod. A large number of fuel rods are disposed surrounding a large diameter water rod and a group of the large-diameter/multi-region poison rods in adjacent with the water rod. The large-diameter water rod has a burnable poison at the tube wall portion. At least a portion of the large-diameter poison rods has a coolant circulation portion allowing coolants to circulate therethrough. Since the large-diameter poison rods are disposed at a position of high neutron fluxes, a large neutron multiplication factor suppression effect can be provided, thereby enabling to reduce the number of burnable poison rods relative to fuels. As a result, power peaking in the fuel assembly is moderated and a greater amount of plutonium can be loaded. In addition the flow of cooling water which tends to gather around the large diameter water rod can be controlled to improve cooling performance of fuels. (N.H.)

  16. Fuel loading and homogeneity analysis of HFIR design fuel plates loaded with uranium silicide fuel

    International Nuclear Information System (INIS)

    Blumenfeld, P.E.

    1995-08-01

    Twelve nuclear reactor fuel plates were analyzed for fuel loading and fuel loading homogeneity by measuring the attenuation of a collimated X-ray beam as it passed through the plates. The plates were identical to those used by the High Flux Isotope Reactor (HFIR) but were loaded with uranium silicide rather than with HFIR's uranium oxide fuel. Systematic deviations from nominal fuel loading were observed as higher loading near the center of the plates and underloading near the radial edges. These deviations were within those allowed by HFIR specifications. The report begins with a brief background on the thermal-hydraulic uncertainty analysis for the Advanced Neutron Source (ANS) Reactor that motivated a statistical description of fuel loading and homogeneity. The body of the report addresses the homogeneity measurement techniques employed, the numerical correction required to account for a difference in fuel types, and the statistical analysis of the resulting data. This statistical analysis pertains to local variation in fuel loading, as well as to ''hot segment'' analysis of narrow axial regions along the plate and ''hot streak'' analysis, the cumulative effect of hot segment loading variation. The data for all twelve plates were compiled and divided into 20 regions for analysis, with each region represented by a mean and a standard deviation to report percent deviation from nominal fuel loading. The central regions of the plates showed mean values of about +3% deviation, while the edge regions showed mean values of about -7% deviation. The data within these regions roughly approximated random samplings from normal distributions, although the chi-square (χ 2 ) test for goodness of fit to normal distributions was not satisfied

  17. Fuel inspection device

    International Nuclear Information System (INIS)

    Tsuji, Tadashi.

    1990-01-01

    The fuel inspection device of the present invention has a feature of obtaining an optimum illumination upon fuel rod interval inspection operation in a fuel pool. That is, an illumination main body used underwater is connected to a cable which is led out on a floor. A light control device is attached to the other end of the cable and an electric power cable is connected to the light control device. A light source (for example, incandescent lamp) is incorporated in the casing of the illumination main body, and a diffusion plate is disposed at the front to provide a plane light source. The light control device has a light control knob capable of remote-controlling the brightness of the light of the illumination main body. In the fuel inspection device thus constituted, halation is scarcely caused on the image screen upon inspection of fuels by a submerged type television camera to facilitate control upon inspection. Accordingly, efficiency of the fuel inspection can be improved to shorten the operation time. (I.S.)

  18. Methodology of ABNT ISO/IEC GUIA 25 implantation in the laboratories of radionuclides analysis in environmental samples of the Analysis Division/CNEN

    International Nuclear Information System (INIS)

    Oliveira, Josue Peter de

    1997-07-01

    The ISO/EEC Guide 25: 1993 Standard G eneral requirements for the competence of calibration and testing laboratories . Is published in Brazil by Brazilian Association for Technical Standards (ABNT) as ABNT ISO/DEC GUIA 25 and establishes general requirements a laboratory must demonstrate to meet, in order to be recognized as having technical competence (accreditation) to carry out specifics calibration or testing. Therefore, the accredited laboratory starts, respectively, taking part from the Brazilian Calibration Network (RBC) or from the Brazilian Testing Laboratories Network (RBLE) . The Environmental Radioanalysis Division (DIAMB) from Environmental Radiological Protection Department (DEPRA) from Institute of Radiation Protection and Dosimetry (IRD) from Brazilian National Nuclear Energy Commission (CNEN) is a laboratory responsible for analyzing radionuclides deriving for the samples from DEPRA's Surveillance Program, research and servings, due to an eventual radionuclide contamination in environment, foods and others raw materials for human consumption; including for importation and exportation products certification purposes. For all these reasons, DIAMB needs its formal recognition for carrying out radionuclides analysis in environmental samples. This work aims to provide a methodology in order to guide a laboratory which has the intention to implement a accreditation process. It also describes policies to meet the requirements related to the Standard, guidance needed to specification of some steps and also comments some points from the Standard in order to become easier all the accreditation process comprehension. (author)

  19. Modern new nuclear fuel characteristics and radiation protection aspects.

    Science.gov (United States)

    Terry, Ian R

    2005-01-01

    The glut of fissile material from reprocessing plants and from the conclusion of the cold war has provided the opportunity to design new fuel types to beneficially dispose of such stocks by generating useful power. Thus, in addition to the normal reactor core complement of enriched uranium fuel assemblies, two other types are available on the world market. These are the ERU (enriched recycled uranium) and the MOX (mixed oxide) fuel assemblies. Framatome ANP produces ERU fuel assemblies by taking feed material from reprocessing facilities and blending this with highly enriched uranium from other sources. MOX fuel assemblies contain plutonium isotopes, thus exploiting the higher neutron yield of the plutonium fission process. This paper describes and evaluates the gamma, spontaneous and alpha reaction neutron source terms of these non-irradiated fuel assembly types by defining their nuclear characteristics. The dose rates which arise from these terms are provided along with an overview of radiation protection aspects for consideration in transporting and delivering such fuel assemblies to power generating utilities.

  20. Inspection of NFT-type cask fabrication

    International Nuclear Information System (INIS)

    Takani, M.; Umegaki, O.

    1998-01-01

    NFT-type cask has been developed to transport the high burn-up spent fuel from Japanese nuclear power stations to the reprocessing plant of Japan Nuclear Fuel Limited which is under construction in Rokkasho-mura, Aomori prefecture. NFT placed orders of 53 casks to 5 fabricators in Japan and overseas, and these casks have been fabricated since 1994. There are two types of NFT-type casks for PWR spent fuel and four types of NFT-type cask for BWR spent fuel. These are designed in consideration of the number of spent fuels accommodated into each type of casks and the handling conditions at domestic nuclear power stations. According to Japanese notification, it is required to be confirmed by competent authority that casks are manufactured in accordance with approved designs. Furthermore, additional tests are performed such as through-gauge test for basket and pressure test on the shielding material space to ensure the performance of cask by NFT other than items inspected by the competent authority. In order to enhance maintainability of casks, replacement parts such as bolts and valves are shared as much as possible. (authors)

  1. Imaging of Flames in Cement Kilns To Study the Influence of Different Fuel Types

    DEFF Research Database (Denmark)

    Pedersen, Morten Nedergaard; Nielsen, Mads; Clausen, Sønnik

    2017-01-01

    The cement industry aims to use an increased amount of alternative fuels to reduce production costs and CO2 emissions. In this study three cement plants firing different kinds and percentages of alternative fuel were studied. A specially developed camera setup was used to monitor the flames...... in the three cement kilns and assess the effect of alternative fuels on the flame. It was found that cofiring with solid recovered fuel (SRF) would delay the ignition point by about 2 m and lower the intensity and temperature of the kiln flame compared to a fossil fuel flame. This is related to a larger...... particle size and moisture content of the alternative fuels, which lowers the conversion rate compared to fossil fuels. The consequences can be a lower kiln temperature and cement quality. The longer conversion time may also lead to the possibility of localized reducing conditions in the cement kiln, which...

  2. Experimental investigation concerning the influence of fuel type and properties on the injection and atomization of liquid biofuels in an optical combustion chamber

    International Nuclear Information System (INIS)

    Galle, J.; Defruyt, S.; Van de Maele, C.; Rodriguez, R. Piloto; Denon, Q.; Verliefde, A.; Verhelst, S.

    2013-01-01

    Due to the scarcity of fossil fuels and the future stringent emission limits, there is an increasing interest for the use of renewable biofuels in compression ignition engines. However, these fuels have different physical, chemical and thermodynamic properties affecting atomization, spray development and combustion processes. The results reported in this paper have been obtained by experimentation with a constant volume combustion chamber. The influences of physical fuel properties on injections under non-evaporating conditions are studied, using a pump-line-nozzle system from a medium speed diesel engine with injection pressures up to 1200 bar, by changing the fuel type and temperature. Experiments were conducted for diesel, biodiesel, straight vegetable oils and animal fats. Injection pressure and needle lift measurements were analyzed. A high speed camera was used to visualize the spray, which enabled us to study the spray penetration and spray angle. Our results show that the fuel temperature is an important parameter to control because it significantly affects the fuel properties. Both the injection timing and injection duration are affected by the fuel properties. The influences of these properties on the spray development were less pronounced. At low temperatures, a strongly deteriorated atomization of oils and fats was observed. -- Highlights: • Spray measurements in an optical combustion chamber. • Influence on the injections system is compared for different bio-fuels. •Temperature effects the fuel properties, with strong influence on the injection system. • Viscosity has significant influence on atomization, especially for viscous fuels. • No difference for spray penetration and angle unlike the mass distribution

  3. Certification test for safety of new fuel transportation package

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Sugawa, Osami; Suga, Masao.

    1993-01-01

    The objective of this certification test is to prove the safety of new fuel transportation package against a fire of actual size caused by traffic accidents. After the fire test, the fuel assemblies were covered with coal-tar like material vaporized from anti-shock material used in the container. Surface color of BWR-type fuel assembly was dark grey that is supposed to be the color of oxide of Zircaloy. As for PWR-type fuel assembly, the condition encountered during fire test caused no change to the outlook of the rod element. Both the BWR and PWR type fuel rod elements showed no deformation and were completely sound. Therefore it may be concluded that the container protected the mimic fuel assemblies against fire of 30 minutes duration and caused no damage. This report is the result of the above experiments and examinations, and we appreciate the cooperation of those who are concerned. (J.P.N.)

  4. Radiation shielding at interim storage facility for CANDU-type nuclear spent fuel

    International Nuclear Information System (INIS)

    Mateescu, S.; Radu, M. Pantazi D.; Stanciu, M.

    1997-01-01

    Technical measures in radiological protection are taken in the interim storage facility design to ensure that, during normal operation, exposures of workers and members of public to ionizing radiation are limited to levels lower than regulatory limits. The spent fuel storage design provides for radiation exposure to be as low as reasonable achievable (ALARA principles). The evaluation of radiation shields includes the most conservative provisions: - all locations which may contain spent fuel are full; - the spent fuel has reached the maximum burnup; - the post irradiation cooling period should be the minimum reasonable; - equipment for handling contains the maximum amount of spent fuel. Radiation shields should ensure that external radiation fields do not exceed limits accepted by the Regulatory Body Module. The evaluation has been performed with two computer codes, QAD-5K and MICROSHIELD-4. (authors)

  5. Polyester synthesis for application in PEMFC type fuel cells; Sintese de poliester para aplicacao em celulas a combustivel do tipo PEM

    Energy Technology Data Exchange (ETDEWEB)

    Fiuza, R.P.; Souza, D.R. de; Fiuza, R.A.; Jose, N.M.; Boaventura, J.S. [Universidade Federal da Bahia (UFBA), Salvador, BA (Brazil). Inst. de Quimica], e-mail: raigenis@gmail.com

    2006-07-01

    The PEMFC (Proton Exchange Membrane Fuel Cell), along the SOFC (Solid Oxide Fuel Cell), is the most important technology, among the various types of fuels cell. The PEMFC shows a large versatility of applications, both for stationary and mobile use. However the PEMFC presents high manufacture cost, directly impacting in the cost of the produced energy. This work contemplates the previews sulfonation of phtalic acid and its subsequent polymerization with glycerol, using as catalytic tin dibutyl-dilaurate. The obtained material has been characterized by DSC, TGA, FTIR, MEV, DRX and XRF. The gotten results indicated that phtalic acid was sulfonated and the increase of the sulfonation degree significantly increased the crystallinity of the sulfonated ftalico acid. Furthermore, the polymer produced from the sulfonated monomer presented adequate thermal resistance and a high content of conducting groups, necessary conditions for application as electrolyte in PEMFC. All these characteristics, particularly the low cost of the reagents and the ease of production process, make the sulfonated polyester membrane a promising candidate as fuel cell electrolyte. (author)

  6. Romanian concern for advanced fuels development

    International Nuclear Information System (INIS)

    Ohai, Dumitru

    2001-01-01

    The Institute for Nuclear Research (ICN), a subsidiary of Romanian Authority for Nuclear Activities, at Pitesti - Romania, has developed a preliminary design of a fuel bundle with 43 elements named SEU 43 for high burnup in CANDU Reactor. A very high experience in nuclear fuels manufacturing and control has also been accumulated. Additionally, on the nuclear site Pitesti there is the Nuclear Fuel Plant (NFP) qualified to manufacturing CANDU 6 type fuel, the main fuel supplier for NPP Cernavoda. A very good collaboration of ICN with NFP can lead to a low cost upgrading the facilities which ensure at present the CANDU standard fuel fabrication to be able of manufacturing also SEU 43 fuel for extended burnup. The financial founds are allocated by Romanian Authority for Nuclear Activities of the Ministry of Industry and Resources to sustain the departmental R and D program 'Nuclear Fuel'. This Program has the main objective to establish a technology for manufacturing a new CANDU fuel type destined for extended burnup. It is studied the possibility to use the Recovered Uranium (RU) resulted from LWR spent fuel reprocessing facility existing in stockpiles. The International Agency for Atomic Energy (IAEA) sustains also this program. By ROM/4/025/ Model Project, IAEA helps ICN to solve the problems regarding materials (RU, Zircaloy 4 tubes) purchasing, devices' upgrading and personnel training. The paper presents the main actions needing to be create the technical base for SEU 43 fuel bundle manufacturing. First step, the technological experiments and experimental fuel element manufacturing, will be accomplished in ICN installations. Second step, the industrial scale, need thorough studies for each installation from NFP to determine tools and technology modification imposed by the new CANDU fuel bundle manufacturing. All modifications must be done such as to the NFP, standard CANDU and SEU fuel bundles to be manufactured alternatively. (author)

  7. Method of transporting fuel assemblies

    International Nuclear Information System (INIS)

    Okada, Katsutoshi.

    1979-01-01

    Purpose: To enable safety transportation of fuel assemblies for FBR type reactors by surrounding each of fuel elements in a wrapper tube by a rubbery, hollow cylindrical container and by sealing medium such as air to the inside of the container. Method: A fuel element is contained in a hollow cylindrical rubber-like tube. The fuel element has an upper end plug, a lower end plug and a wire spirally wound around the outer periphery. Upon transportation of the fuel assemblies, each of the fuel elements is covered with the container and arranged in the wrapper tube and then the fuel assemblies are assembled. Then, medium such as air is sealed for each of the fuel elements by way of an opening and then the opening is tightly closed. Before loading the transported fuel assemblies in the reactor, the medium is discharged through the opening and the container is completely extracted and removed from the inside of the wrapper tube. (Seki, T.)

  8. Laboratory measurements of trace gas emissions from biomass burning of fuel types from the southeastern and southwestern United States

    Directory of Open Access Journals (Sweden)

    I. R. Burling

    2010-11-01

    Full Text Available Vegetation commonly managed by prescribed burning was collected from five southeastern and southwestern US military bases and burned under controlled conditions at the US Forest Service Fire Sciences Laboratory in Missoula, Montana. The smoke emissions were measured with a large suite of state-of-the-art instrumentation including an open-path Fourier transform infrared (OP-FTIR spectrometer for measurement of gas-phase species. The OP-FTIR detected and quantified 19 gas-phase species in these fires: CO2, CO, CH4, C2H2, C2H4, C3H6, HCHO, HCOOH, CH3OH, CH3COOH, furan, H2O, NO, NO2, HONO, NH3, HCN, HCl, and SO2. Emission factors for these species are presented for each vegetation type burned. Gas-phase nitrous acid (HONO, an important OH precursor, was detected in the smoke from all fires. The HONO emission factors ranged from 0.15 to 0.60 g kg−1 and were higher for the southeastern fuels. The fire-integrated molar emission ratios of HONO (relative to NOx ranged from approximately 0.03 to 0.20, with higher values also observed for the southeastern fuels. The majority of non-methane organic compound (NMOC emissions detected by OP-FTIR were oxygenated volatile organic compounds (OVOCs with the total identified OVOC emissions constituting 61 ± 12% of the total measured NMOC on a molar basis. These OVOC may undergo photolysis or further oxidation contributing to ozone formation. Elevated amounts of gas-phase HCl and SO2 were also detected during flaming combustion, with the amounts varying greatly depending on location and vegetation type. The fuels with the highest HCl emission factors were all located in the coastal regions, although HCl was also observed from fuels farther inland. Emission factors for HCl were generally higher for the southwestern fuels

  9. The modeling experience of fuel element units operation under MSC.MARC and MENTAT 2008R1

    International Nuclear Information System (INIS)

    Kulakov, G.; Kashirin, B.; Kosaurov, A.; Konovalov, Y.; Kuznetsov, A.; Medvedev, A.; Novikov, V.; Vatulin, A.

    2009-01-01

    MSC Software is leading developer of CAE-software in the world, so behaviour of fuel elements modeling with MSC.MARC use is of great practical importance. Behaviour of fuel elements usually is modeled in the elastic-viscous-plastic statement with account on fuel swelling during irradiation. For container type fuel elements contact interaction between fuel pellets and cladding or other parts of fuel element in top and bottom plugs must be in account. Results of simulated behaviour of various type fuel elements - container type fuel elements for PWR and RBMK reactors, dispersion type fuel elements for research reactors are presented. (authors)

  10. Spent fuel transport in fuel cycle

    International Nuclear Information System (INIS)

    Labrousse, M.

    1977-01-01

    The transport of radioactive substances is a minor part of the fuel cycle because the quantities of matter involved are very small. However the length and complexity of the cycle, the weight of the packing, the respective distances between stations, enrichment plants and reprocessing plants are such that the problem is not negligible. In addition these transports have considerable psychological importance. The most interesting is spent fuel transport which requires exceptionally efficient packaging, especially where thermal and mechanical resistance are concerned. To meet the safety criteria necessary for the protection of both public and users it was decided to use the maximum capacity consistent with rail transport and to avoid coolant fluids under pressure. Since no single type of packing is suitable for all existing stations an effort has been made to standardise handling accessories, and future trands are towards maximum automation. A discussion on the various technical solutions available for the construction of these packing systems is followed by a description of those used for the two types of packaging ordered by COGEMA [fr

  11. Sensitivity Analysis of Criticality for Different Nuclear Fuel Shapes

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Hyun Sik; Jang, Misuk; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)

    2016-10-15

    Rod-type nuclear fuel was mainly developed in the past, but recent study has been extended to plate-type nuclear fuel. Therefore, this paper reviews the sensitivity of criticality according to different shapes of nuclear fuel types. Criticality analysis was performed using MCNP5. MCNP5 is well-known Monte Carlo codes for criticality analysis and a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical systems. We performed the sensitivity analysis of criticality for different fuel shapes. In sensitivity analysis for simple fuel shapes, the criticality is proportional to the surface area. But for fuel Assembly types, it is not proportional to the surface area. In sensitivity analysis for intervals between plates, the criticality is greater as the interval increases, but if the interval is greater than 8mm, it showed an opposite trend that the criticality decrease by a larger interval. As a result, it has failed to obtain the logical content to be described in common for all cases. The sensitivity analysis of Criticality would be always required whenever subject to be analyzed is changed.

  12. Sensitivity Analysis of Criticality for Different Nuclear Fuel Shapes

    International Nuclear Information System (INIS)

    Kang, Hyun Sik; Jang, Misuk; Kim, Seoung Rae

    2016-01-01

    Rod-type nuclear fuel was mainly developed in the past, but recent study has been extended to plate-type nuclear fuel. Therefore, this paper reviews the sensitivity of criticality according to different shapes of nuclear fuel types. Criticality analysis was performed using MCNP5. MCNP5 is well-known Monte Carlo codes for criticality analysis and a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical systems. We performed the sensitivity analysis of criticality for different fuel shapes. In sensitivity analysis for simple fuel shapes, the criticality is proportional to the surface area. But for fuel Assembly types, it is not proportional to the surface area. In sensitivity analysis for intervals between plates, the criticality is greater as the interval increases, but if the interval is greater than 8mm, it showed an opposite trend that the criticality decrease by a larger interval. As a result, it has failed to obtain the logical content to be described in common for all cases. The sensitivity analysis of Criticality would be always required whenever subject to be analyzed is changed

  13. Fuels and Combustion

    KAUST Repository

    Johansson, Bengt

    2016-08-17

    This chapter discusses the combustion processes and the link to the fuel properties that are suitable for them. It describes the basic three concepts, including spark ignition (SI) and compression ignition (CI), and homogeneous charge compression ignition (HCCI). The fuel used in a CI engine is vastly different from that in an SI engine. In an SI engine, the fuel should sustain high pressure and temperature without autoignition. Apart from the dominating SI and CI engines, it is also possible to operate with a type of combustion: autoignition. With HCCI, the fuel and air are fully premixed before combustion as in the SI engine, but combustion is started by the increased pressure and temperature during the compression stroke. Apart from the three combustion processes, there are also a few combined or intermediate concepts, such as Spark-Assisted Compression Ignition (SACI). Those concepts are discussed in terms of the requirements of fuel properties.

  14. Fuels and Combustion

    KAUST Repository

    Johansson, Bengt

    2016-01-01

    This chapter discusses the combustion processes and the link to the fuel properties that are suitable for them. It describes the basic three concepts, including spark ignition (SI) and compression ignition (CI), and homogeneous charge compression ignition (HCCI). The fuel used in a CI engine is vastly different from that in an SI engine. In an SI engine, the fuel should sustain high pressure and temperature without autoignition. Apart from the dominating SI and CI engines, it is also possible to operate with a type of combustion: autoignition. With HCCI, the fuel and air are fully premixed before combustion as in the SI engine, but combustion is started by the increased pressure and temperature during the compression stroke. Apart from the three combustion processes, there are also a few combined or intermediate concepts, such as Spark-Assisted Compression Ignition (SACI). Those concepts are discussed in terms of the requirements of fuel properties.

  15. Evaluation of analysis method standardless by WDXRF and EDXRF of aluminum powder used in MTR type fuel

    International Nuclear Information System (INIS)

    Scapin, Valdirene O.; Salvador, Vera L.R.; Cotrim, Marycel E.B.; Pires, Maria A.F.; Scapin, Marcos A.

    2011-01-01

    The nuclear fuel used in IEA-R1m reactor at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP) is the MTR type. This fuel is compound of a core (U 3 Si 2 -Al dispersion briquette) wrapped in an aluminum plate with two cladding (superior and inferior) both in aluminum. The fuel element efficiency depends on the quality control of U 3 Si 2 and aluminum. For aluminum should be checked the impurities levels such as Si, Mn, Fe, Co, Cu, Zn and others and Al total . Aiming to provide a quick method, multielemental and non-destructive, the performance of the wavelength dispersive (WDXRF) and energy dispersive (EDXRF) X-ray fluorescence techniques, using the curve instrument sensitivity curve method, also known like standard less analysis, was evaluated. This method allows the determination from the element boron (Z=5) to uranium (Z=92) with concentrations ranging from 0.001 to 99.99% without the need for individual calibration curve and chemical pretreatments in the sample preparation. The results were compared with calibration curve method data, using statistical tests tools. By multivariate analysis of all the experimental data, especially by the discriminant analysis (DA) and cluster analysis (CA), respectively, it was possible to evaluate a correlation between variables of the applied analytical methods could be interpreted in context to qualify the fuels by XRF technique and method standard less. The results showed that the proposed method is satisfactory for both spectrometers; however it was found that the WDXRF presents the greatest conformity degree. (author)

  16. Storage of Spent Nuclear Fuel. Specific Safety Guide

    International Nuclear Information System (INIS)

    2012-01-01

    This Safety Guide provides recommendations and guidance on the storage of spent nuclear fuel. It covers all types of storage facilities and all types of spent fuel from nuclear power plants and research reactors. It takes into consideration the longer storage periods that have become necessary owing to delays in the development of disposal facilities and the decrease in reprocessing activities. It also considers developments associated with nuclear fuel, such as higher enrichment, mixed oxide fuels and higher burnup. The Safety Guide is not intended to cover the storage of spent fuel if this is part of the operation of a nuclear power plant or spent fuel reprocessing facility. Guidance is provided on all stages for spent fuel storage facilities, from planning through siting and design to operation and decommissioning, and in particular retrieval of spent fuel. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. Management system; 5. Safety case and safety assessment; 6. General safety considerations for storage of spent fuel. Appendix I: Specific safety considerations for wet or dry storage of spent fuel; Appendix II: Conditions for specific types of fuel and additional considerations; Annex: I: Short term and long term storage; Annex II: Operational and safety considerations for wet and dry spent fuel storage facilities; Annex III: Examples of sections of operating procedures for a spent fuel storage facility; Annex IV: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex V: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex VI: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex VII: Postulated initiating events for consideration in a safety assessment (internal phenomena).

  17. Hydrogen fueling stations in Japan hydrogen and fuel cell demonstration project

    International Nuclear Information System (INIS)

    Koseki, K.; Tomuro, J.; Sato, H.; Maruyama, S.

    2004-01-01

    A new national demonstration project of fuel cell vehicles, which is called Japan Hydrogen and Fuel Cell Demonstration Project (JHFC Project), has started in FY2002 on a four-year plan. In this new project, ten hydrogen fueling stations have been constructed in Tokyo and Kanagawa area in FY2002-2003. The ten stations adopt the following different types of fuel and fueling methods: LPG reforming, methanol reforming, naphtha reforming, desulfurized-gasoline reforming, kerosene reforming, natural gas reforming, water electrolysis, liquid hydrogen, by-product hydrogen, and commercially available cylinder hydrogen. Approximately fifty fuel cell passenger cars and a fuel cell bus are running on public roads using these stations. In addition, two hydrogen stations will be constructed in FY2004 in Aichi prefecture where The 2005 World Exposition (EXPO 2005) will be held. The stations will service eight fuel cell buses used as pick-up buses for visitors. We, Engineering Advancement Association of Japan (ENAA), are commissioned to construct and operate a total of twelve stations by Ministry of Economy Trade and Industry (METI). We are executing to demonstrate or identify the energy-saving effect, reduction of the environmental footprint, and issues for facilitating the acceptance of hydrogen stations on the basis of the data obtained from the operation of the stations. (author)

  18. Development of quality assurance methods for low enriched fuel assemblies

    International Nuclear Information System (INIS)

    Woolstenhulme, N.E.; Moore, G.A.; Perez, D.M.; Wachs, D.M.

    2010-01-01

    As the Reduced Enrichment for Research and Test Reactors (RERTR) fuel development program has furthered the technology of low enriched uranium fuels, much effort has been expended to specify requirements, perform appropriate inspections, and to qualify experimental fuel plates and assemblies for irradiation. A great deal of consideration has been given to generate examinations and criteria that are both applicable to the unique fuel types being developed and consistent with industry practices for inspecting plate-type reactor fuel. Recent developments in quality assurance (QA) methodologies have given a heightened confidence in satisfactory fuel plate performance. At the same time, recommendations are given to further develop a system suitable for the testing and acceptance of production fuel elements containing low enriched uranium fuels. (author)

  19. Westinghouse introduces new fuel for PWRs and BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Orr, W L; McClintock, D C

    1985-09-01

    In response to utility demands for improved fuel performance, reduced fuel cycle costs, and enhanced operating margins, Westinghouse recently introduced advanced fuel assembly designs for both types of LWR - Vantage 5 for PWRs, and Quad+ for BWRs.

  20. Analysis of fuel pin behavior under slow-ramp type transient overpower condition by using the fuel performance evaluation code 'FEMAXI-FBR'

    International Nuclear Information System (INIS)

    Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

    2012-01-01

    FEMAXI-FBR has been developed as the one module of the core disruptive accident analysis code 'ASTERIA-FBR' in order to evaluate the mixed oxide (MOX) fuel performance under steady, transient and accident conditions of fast reactors consistently. On the basis of light water reactor (LWR) fuel performance evaluation code 'FEMAXI-6', FEMAXI-FBR develops specific models for the fast reactor fuel performance, such as restructuring, material migration during steady state and transient, melting cavity formation and pressure during accident, so that it can evaluate the fuel failure during accident. The analysis of test pin with slow transient over power test of CABRI-2 program was conducted from steady to transient. The test pin was pre-irradiated and tested under transient overpower with several % P 0 /s (P 0 : steady state power) of the power rate. Analysis results of the gas release ratio, pin failure time, and fuel melt radius were compared to measured values. The analysis results of the steady and transient performances were also compared with the measured values. The compared performances are gas release ratio, fuel restructuring for steady state and linear power and melt radius at failure during transient. This analysis result reproduces the measured value. It was concluded that FEMAXI-FBR is effective to evaluate fast reactor fuel performances from steady state to accident conditions. (author)

Some links on this page may take you to non-federal websites. Their policies may differ from this site.