Ondas de choque em jatos de quasares e objetos BL Lacertae

NASA Astrophysics Data System (ADS)

Melo, F. E.; Botti, L. C. L.

2003-08-01

Este trabalho é parte de um projeto que vem sendo realizado há dois anos no CRAAM, cujos objetivos principais são analisar e aplicar um modelo generalizado de ondas de choque em jatos relativísticos de plasma, presentes em quasares e objetos BL Lacertae, para explicar a variabilidade observada nestes objetos. O método consiste em uma decomposição de curvas de luz em séries de explosões similares, em várias freqüências, baseando-se em uma evolução espectro-temporal média das explosões. A partir da evolução média, um ajuste de cada explosão é feito com base em equações empíricas, modificando-se apenas parâmetros específicos de cada explosão. Inicialmente o modelo foi aplicado ajustando-se as curvas de luz a explosões delineadas por uma evolução do choque em três estágios, segundo a predominância do processo de emissão: síncrotron, Compton e adiabático. Entretanto, nesta nova fase de projeto, visando uma parametrização mais concisa, uma otimização do algoritmo de ajuste e uma convergência mais rápida, a formulação para cada evento foi assumida com uma evolução em apenas dois estágios: subida e descida. Isto possibilitou uma ótima delineação das curvas de luz das fontes OV236, OJ287, 3C273 e BL Lac, entre 1980 e 2000, nas freqüências 4.8, 8.0, 14.5 e 22 GHz, utilizando-se dados do Observatório da Universidade de Michigan, do Observatório do Itapetinga (Atibaia SP) e do Observatório Metsähovi. Como conclusões importantes, verificou-se que: os parâmetros ajustados descrevem o comportamento do jato; os valores do índice que descreve a expansão do jato sugerem que o mesmo se expande de uma forma não-cônica; o campo magnético é turbulento atrás da frente de choque; e as peculiaridades das explosões são devidas à influência de grandezas tais como o coeficiente da distribuição espectral de energia dos elétrons, a intensidade de campo magnético e o fator de feixe Doppler, no início do choque.

Development and validation of the Hospitality Axiological Scale for Humanization of Nursing Care.

PubMed

Galán González-Serna, José María; Ferreras-Mencia, Soledad; Arribas-Marín, Juan Manuel

2017-08-03

to develop and validate a scale to evaluate nursing attitudes in relation to hospitality for the humanization of nursing care. Participants: the sample consisted of 499 nursing professionals and undergraduate students of the final two years of the Bachelor of Science in Nursing program. the instrument has been developed and validated to evaluate the ethical values related to hospitality using a methodological approach. Subsequently, a model was developed to measure the dimensions forming the construct hospitality. the Axiological Hospitality Scale showed a high internal consistency, with Cronbach's Alpha=0.901. The validation of the measuring instrument was performed using factorial, exploratory and confirmatory analysis techniques with high goodness of fit measures. the developed instrument showed an adequate validity and a high internal consistency. Based on the consistency of its psychometric properties, it is possible to affirm that the scale provides a reliable measurement of the hospitality. It was also possible to determine the dimensions or sources that embrace it: respect, responsibility, quality and transpersonal care. desenvolver e validar uma escala que permita avaliar a atitude dos enfermeiros em termos de hospitalidade, visando a humanização da enfermagem.Participantes: a amostra foi constituída por 499 profissionais e estudantes de enfermagem dos dois últimos anos do curso de graduação em Enfermagem. utilizando-seuma abordagem metodológica, foi desenvolvido e validado um instrumento para avaliar os valores éticos relacionados com a hospitalidade. Subsequentemente, foi formulado um modelo para mediras dimensões que constituem o construto hospitalidade. a Escala Axiológica de Hospitalidade mostrou uma consistência interna elevada, com Alfa de Cronbach=0,901. A validação do instrumento de medição foi realizada usando-se métodos de análise fatorial, exploratória e confirmatória, que apresentaram bons índices de qualidade de ajuste. o instrumento desenvolvido apresentou uma validade adequada e uma consistência interna elevada. Com base na consistência de suas propriedades psicométricas, é possível afirmar que a escala proporciona uma medida confiável da hospitalidade. Também foi possível determinar as dimensões ou fontes que a compõem: o respeito, a responsabilidade, a qualidade e o cuidado transpessoal. desarrollar y validar una escala que permita evaluar la actitud enfermera ante la hospitalidad para la humanización de enfermería. Participantes: la muestra la componen 499 profesionales de enfermería y estudiantes de los dos últimos niveles académicos de la titulación de Grado en Enfermería. mediante abordaje metodológico, se ha desarrollado y validado un instrumento para evaluar valores éticos relacionados con la hospitalidad. Posteriormente, se ha formulado un modelo de medida de las dimensiones que conforman el constructo hospitalidad. la Escala Axiológica de la Hospitalidad ha mostrado una alta consistencia interna, con un Alfa de Cronbach=0.901. La validación del instrumento de medida se realizó mediante las técnicas de análisis factorial, exploratorio y confirmatorio, que presentaron unos buenos índices de bondad de ajuste. el instrumento desarrollado ha presentado una adecuada validez y una alta consistencia interna. Sus adecuadas propiedades psicométricas permiten afirmar que la escala aporta una medición fiable de la hospitalidad. También ha permitido determinar las dimensiones o fuentes que la conforman: el respeto, la responsabilidad, la calidad y el cuidado transpersonal.

Clinical Validation of the Nursing Outcome "Swallowing Status" in People with Stroke: Analysis According to the Classical and Item Response Theories.

PubMed

Oliveira-Kumakura, Ana Railka de Souza; de Araujo, Thelma Leite; Costa, Alice Gabrielle de Sousa; Cavalcante, Tahissa Frota; Lopes, Marcos Venícios de Oliveira; Carvalho, Emilia Campos

2017-09-19

To validate clinically the nursing outcome "Swallowing status". The adjustment of the nursing outcome was investigated according to the Classical and Item Response Theories. The models were compared regarding information loss, goodness-of-fit, and differential item functioning. Stability and internal consistency were examined. The nursing outcome has the best fit in the generalized partial credit model with different discrimination parameters. Strong correlations among the scores of each indicator were observed. There was no differential item functioning of the outcome indicators. The scale presented high internal consistency (Cronbach's α = .954) and stability (and > .800). This study presents a valid nursing outcome. Most accurate monitoring of sensitivity to an intervention. Validar clinicamente o resultado de enefermagem "Estado da Deglutição". MÉTODOS: O ajustamento do resultado foi investigado de acordo com as teorias Clássica e de Resposta ao Item. Os modelos foram comparados assumindo parâmetros de itens cruzados de igual discriminação. Investigaram-se as propriedades de bondade do ajuste, funcionamento diferencial dos itens, estabilidade e consistência interna. O resultado se ajustou melhor a partir do Modelo de crédito parcial generalizado, o qual demonstrou unidimensionalidade do resultado e forte correlação entre os escores de cada indicador. Não houve funcionamento diferencial dos indicadores. A consistência interna para a escala global (Cronbach's α = .954) e a estabilidade (>.800) mantiveram-se elevadas. CONCLUSÃO: O estudo apresenta um resultado de enfermagem válido. RELEVÂNCIA PARA A PRÁTICA CLÍNICA: Maior acurácia para monitorar a sensibilidade da intervenção. © 2017 NANDA International, Inc.

Datação do disco galáctico pela nucleocosmocronologia do [Th/Eu

NASA Astrophysics Data System (ADS)

del Peloso, E. F.; da Silva, L.; Arany-Prado, L. I.

2003-08-01

A nucleocosmocronologia emprega abundâncias de nuclídeos radioativos na datação de escalas de tempo astrofísicas. O 232Th é um nuclídeo radioativo com meia-vida de 14 Gano, enquanto que os dois isótopos mais abundantes do Eu são estáveis. O decaimento radioativo do Th modifica as razões de abundâncias [Th/Eu], fornecendo assim um meio de sondar a escala de formação das populações estelares. O objetivo deste trabalho é averiguar a possibilidade de estimar uma idade para o disco Galáctico através da nucleocosmocronologia do [Th/Eu] e investigar o nível de incerteza associado a esta estimativa. Para tanto, foi selecionada uma amostra de 20 estrelas anãs ou subgigantes de tipos espectrais F5 a G9, com -1,00 £ [Fe/H] £ +0,30 e idade(Gano) £ 13. As abundâncias de Th e Eu foram obtidas por síntese espectral das linhas localizadas em 4019,1 Å e 4129,7 Å, respectivamente. Uma comparação destas abundâncias com outros resultados da literatura demonstra que nossos valores apresentam dispersão 2 a 3 vezes menor que qualquer trabalho anterior. Os parâmetros atmosféricos e abundâncias dos elementos que contaminam as regiões espectrais destas linhas foram determinados por nós, de maneira totalmente autoconsistente, através de análise espectral detalhada diferencial em relação ao Sol. As idades estelares individuais foram determinadas através de curvas isócronas teóricas no diagrama HR. Foi realizada, então, uma análise cronológica dos gráficos [Th/Eu] vs. [Fe/H] e [Th/Eu] vs. idade. Os dados estelares foram comparados a curvas calculadas para 3 idades do disco Galáctico - 9, 12, 15 Gano - e foi estudada a sensibilidade à idade assumida no cálculo do ajuste destas curvas aos dados. Estas curvas foram calculadas com base num modelo analítico de evolução química da Galáxia que leva em consideração a formação de refugos, que são compostos pelos remanescentes da evolução estelar, pelos resíduos da formação de estrelas de baixa massa (planetas, cometas, etc.) e por quaisquer outros objetos de massa não-estelar. A formação de resíduos tem o efeito indireto de diluir o meio interestelar, levando a um enriquecimento mais lento deste e a um bom ajuste de diversos vínculos da evolução química da Galáxia, como a distribuição de anãs-G e a relação idade-metalicidade. Os efeitos da destruição do Th por reações fotonucleares em interiores estelares também foram considerados.

Bioreduction of para-chloronitrobenzene in a hydrogen-based hollow-fiber membrane biofilm reactor: effects of nitrate and sulfate.

PubMed

Li, Haixiang; Zhang, Zhiqiang; Xu, Xiaoyin; Liang, Jun; Xia, Siqing

2014-04-01

A continuous-stirred, hydrogen-based, hollow-fiber membrane biofilm reactor (HFMBfR) that was active in nitrate and sulfate reductions was shown to be effective for degradation or detoxification of para-chloronitrobenzene (p-CNB) in water by biotransforming it first to para-chloroaniline (nitro-reduction) and then to aniline (reductive dechlorination) with hydrogen (H2) as an electron donor. A series of short-term experiments examined the effects of nitrate and sulfate on p-CNB bioreduction. The results obtained showed both higher nitrate and sulfate concentration declined the p-CNB bioreduction in the biofilm, and this suggests the competition for H2 caused less H2 available for the p-CNB bioreduction when the H2 demand for the reductions was larger. Denitrification and sulfate reduction intermediates were thought to be potential factors inhibiting the p-CNB bioreduction. Analysis of electron-equivalent fluxes and reaction orders in the biofilm further demonstrated both denitrification and sulfate reduction competed more strongly for H2 availability than p-CNB bioreduction. These findings have significant implications for the HFMBfR used for degrading p-CNB under denitrifying and/or sulfate reducing conditions.

Kinetics of thermophilic anaerobes in fixed-bed reactors.

PubMed

Perez, M; Romero, L I; Sales, D

2001-08-01

The main objective of this study is to estimate growth kinetic constants and the concentration of "active" attached biomass in two anaerobic thermophilic reactors which contain different initial sizes of immobilized anaerobic mixed cultures and decompose distillery wastewater. This paper studies the substrate decomposition in two lab-scale fixed-bed reactors operating at batch conditions with corrugated tubes as support media. It can be demonstrated that high micro-organisms-substrate ratios favor the degradation activity of the different anaerobic cultures, allowing the stable operation without lag-phases and giving better quality in effluent. The kinetic parameters obtained--maximum specific growth rates (mu(max)), non-biodegradable substrate (S(NB)) and "active or viable biomass" concentrations (X(V0))--were obtained by applying the Romero kinetic model [L.I. Romero, 1991. Desarrollo de un modelo matemático general para los procesos fermentativos, Cinética de la degradación anaerobia, Ph.D. Thesis, University of Cádiz (Spain), Serv. Pub. Univ. Cádiz], with COD as substrate and methane (CH4) as the main product of the anaerobic process. This method is suitable to calculate and to differentiate the main kinetic parameters of both the total anaerobic mixed culture and the methanogenic population. Comparison of experimental measured concentration of volatile attached solids (VS(att)) in both reactors with the estimated "active" biomass concentrations obtained by applying Romero kinetic model [L.I. Romero, 1991. Desarrollo de un modelo matemático general para los procesos fermentativos, Cinética de la degradación anaerobia, Ph.D. Thesis, University of Cádiz (Spain), Serv. Pub. Univ. Cádiz] shows that a large amount of inert matter is present in the fixed-bed reactor.

Accidentes laborales asociados al desánimo de médicos SERUMS para laborar en el primer nivel de atención de Lima, Perú

PubMed Central

Mejia, Christian R.; Valladares-Garrido, Mario J.; Romero, Brian M.; Valladares-Garrido, Danai; Linares-Reyes, Edgardo

2018-01-01

Introducción la retención laboral es un tema de suma importancia, porque se requiere de profesionales en el primer nivel de atención (PNA). El objetivo fue determinar si los accidentes laborales se asociaron al desánimo de los médicos para trabajar en el PNA de Lima, Perú. Métodos estudio transversal analítico de datos secundarios de una base de datos generada de una encuesta a médicos que realizaron su Servicio Rural y Urbano-Marginal en Salud (SERUMS). Se incluyó solo a los médicos que manifestaron al inicio del SERUMS que podían trabajar en el PNA de Lima. Se definió como cambio de intención de trabajo en el PNA a los que finalizando el SERUMS refirieron que ya no deseaban laborar en Lima. Esto se asoció según si tuvieron un accidente laboral y se ajustó por otras variables. Resultados de los 124 médicos el 63% fueron hombres (78). La mediana de edad fue de 26 años (rango intercuartílico: 25-27 años). Después de su SERUMS, el 12% (15) manifestó que cambió su interés y que deseaba trabajar en la capital. En el análisis multivariado, haber tenido un accidente laboral disminuyó la frecuencia del cambio de intención de trabajo en el PNA (RPa: 0.28, IC 95%: 0.14-0.54, p < 0,001), ajustado por ocho variables. Conclusiones en un estudio previo los accidentes laborales disminuyeron la frecuencia de trabajar en provincias, pero nuestro estudio dice lo contrario, posiblemente por la percepción de que un trabajo en la capital del país permite estar más cerca de los servicios para ser atendido en caso de cualquier emergencia. PMID:29190859

Investigation of a para-ortho hydrogen reactor for application to spacecraft sensor cooling

NASA Technical Reports Server (NTRS)

Nast, T. C.

1983-01-01

The utilization of solid hydrogen in space for sensor and instrument cooling is a very efficient technique for long term cooling or for cooling at high heat rates. The solid hydrogen can provide temperatures as low as 7 to 8 K to instruments. Vapor cooling is utilized to reduce parasitic heat inputs to the 7 to 8 K stage and is effective in providing intermediate cooling for instrument components operating at higher temperatures. The use of solid hydrogen in place of helium may lead to weight reductions as large as a factor of ten and an attendent reduction in system volume. The results of an investigation of a catalytic reactor for use with a solid hydrogen cooling system is presented. Trade studies were performed on several configurations of reactor to meet the requirements of high reactor efficiency with low pressure drop. Results for the selected reactor design are presented for both liquid hydrogen systems operating at near atmospheric pressure and the solid hydrogen cooler operating as low as 1 torr.

Continuous hyperpolarization with parahydrogen in a membrane reactor

NASA Astrophysics Data System (ADS)

Lehmkuhl, Sören; Wiese, Martin; Schubert, Lukas; Held, Mathias; Küppers, Markus; Wessling, Matthias; Blümich, Bernhard

2018-06-01

Hyperpolarization methods entail a high potential to boost the sensitivity of NMR. Even though the "Signal Amplification by Reversible Exchange" (SABRE) approach uses para-enriched hydrogen, p-H2, to repeatedly achieve high polarization levels on target molecules without altering their chemical structure, such studies are often limited to batch experiments in NMR tubes. Alternatively, this work introduces a continuous flow setup including a membrane reactor for the p-H2, supply and consecutive detection in a 1 T NMR spectrometer. Two SABRE substrates pyridine and nicotinamide were hyperpolarized, and more than 1000-fold signal enhancement was found. Our strategy combines low-field NMR spectrometry and a membrane flow reactor. This enables precise control of the experimental conditions such as liquid and gas pressures, and volume flow for ensuring repeatable maximum polarization.

Fotometria superficial BVRI de 18 galáxias fracas

NASA Astrophysics Data System (ADS)

Saraiva, M. F. O.; Silva, P. R.

2003-08-01

Conhecer as propriedades de galáxias a diferentes redshifts é uma questão fundamental para entender o problema da formação e evolução das galáxias, e desde a década passada tem se intensificado fortemente o estudo de galáxias muito distantes. No entanto parece haver um interesse menor em galáxias a distâncias intermediárias, que aparecem como objetos de fundo em imagens de objetos próximos, e que são igualmente importantes. Examinando imagens BVRI de longa exposição, ótimo sinal/ruído, grande campo (46'x46'), das vizinhanças de NGC 7479, detectamos 18 galáxias fracas (18 < B < 21) nessas imagens. Neste trabalho, apresentamos a fotometria superficial desses objetos. Determinamos coordenadas equatoriais, magnitudes e cores integradas, perfis de brilho e de cor, e parâmetros isofotais calculados por ajuste de ellipses, dentro do limite permitido pela baixa resolução espacial dos dados (1,35 segarc/pixel). Nosso objetivo é procurar correlações entre as propriedades das galáxias e, tentativamente, comparar esses dados com aqueles de galáxias de redshift conhecido, disponíveis na literatura, para estimar suas distâncias (a partir da relação cor x redshift) e suas morfologias. Uma análise preliminar nesse sentido mostrou que as cores aparentes B-V, V-R e V-I dos objetos da nossa amostra, a menos de duas exceções, ocupam regiões bem definidas nos diagramas cor-cor, e não apresentam diferenças notáveis em relação às cores típicas de galáxias próximas.

Four Quarters of Football Helmet Safety

MedlinePlus

... Association). 082012 Pub. 348 AJUSTE CORRECTO UN JUEGO MÁS SEGURO 4 CUARTOS SOBRE LA SEGURIDAD DE LOS ... La salud a largo plazo del atleta es más importante que el resultado de un juego. Recuerde ...

Bioreduction of para-chloronitrobenzene in drinking water using a continuous stirred hydrogen-based hollow fiber membrane biofilm reactor.

PubMed

Xia, Siqing; Li, Haixiang; Zhang, Zhiqiang; Zhang, Yanhao; Yang, Xin; Jia, Renyong; Xie, Kang; Xu, Xiaotian

2011-08-30

para-Chloronitrobenzene (p-CNB) is particularly harmful and persistent in the environment and is one of the priority pollutants. A feasible degradation pathway for p-CNB is bioreduction under anaerobic conditions. Bioreduction of p-CNB using a hydrogen-based hollow fiber membrane biofilm reactor (HFMBfR) was investigated in the present study. The experiment results revealed that p-CNB was firstly reduced to para-chloraniline (p-CAN) as an intermediate and then reduced to aniline that involves nitro reduction and reductive dechlorination with H(2) as the electron donor. The HFMBfR had reduced p-CNB to a major extent with a maximum removal percentage of 99.3% at an influent p-CNB concentration of 2mg/L and a hydraulic residence time of 4.8h, which corresponded to a p-CNB flux of 0.058g/m(2) d. The H(2) availability, p-CNB loading, and the presence of competing electron acceptors affected the p-CNB reduction. Flux analysis indicated that the reduction of p-CNB and p-CAN could consume fewer electrons than that of nitrate and sulfate. The HFMBfR had high average hydrogen utilization efficiencies at different steady states in this experiment, with a maximum efficiency at 98.2%. Copyright © 2011 Elsevier B.V. All rights reserved.

Supported transition metal catalysts for para- to ortho-hydrogen conversion

NASA Technical Reports Server (NTRS)

Brooks, Christopher J.; Wang, Wei; Eyman, Darrell P.

1994-01-01

The main goal of this study was to develop and improve on existing catalysts for the conversion of ortho- to para-hydrogen. Starting with a commercially available Air Products nickel silicate, which had a beta value of 20, we were trying to synthesize catalysts that would be an improvement to AP. This was accomplished by preparing silicates with various metals as well as different preparation methods. We also prepared supported ruthenium catalysts by various techniques using several metal precursors to improve present technology. What was also found was that the activation conditions prior to catalytic testing was highly important for both the silicates and the supported ruthenium catalysts. While not the initial focus of the research, we made some interesting observations into the adsorption of H2 on ruthenium. This helped us to get a better understanding of how ortho- to para-H2 conversion takes place, and what features in a catalyst are important to optimize activity. Reactor design was the final area in which some interesting conclusions were drawn. As discussed earlier, the reactor catalyst bed must be constructed using straight 1/8 feet OD stainless steel tubing. It was determined that the use of 1/4 feet OD tubing caused two problems. First, the radius from the center of the bed to the wall was too great for thermal equilibrium. Since the reaction of ortho- to para-H2 is exothermic, the catalyst bed center was warmer than the edges. Second, the catalyst bed was too shallow using a 1/4 feet tube. This caused reactant blow-by which was thought to decrease the measured activity when the flow rate was increased. The 1/8 feet tube corrected both of these concerns.

Distribution of stable isotopes in arid storms . II. A double-component model of kinematic wave flow and transport

NASA Astrophysics Data System (ADS)

Yakirevich, Alexander; Dody, Avraham; Adar, Eilon M.; Borisov, Viacheslav; Geyh, Mebus

A new mathematical method based on a double-component model of kinematic wave flow and approach assesses the dynamic isotopic distribution in arid rain storms and runoff. This model describes the transport and δ18O evolution of rainfall to overland flow and runoff in an arid rocky watershed with uniformly distributed shallow depression storage. The problem was solved numerically. The model was calibrated using a set of temporal discharge and δ18O distribution data for rainfall and runoff collected on a small rocky watershed at the Sede Boker Experimental Site, Israel. Simulation of a reliable result with respect to observation was obtained after parameter adjustment by trial and error. Sensitivity analysis and model application were performed. The model is sensitive to changes in parameters characterizing the depression storage zones. The model reflects the effect of the isotopic memory in the water within the depression storage between sequential rain spells. The use of the double-component model of kinematic wave flow and transport provides an appropriate qualitative and quantitative fitting between computed and observed δ18O distribution in runoff. RésuméUne nouvelle méthode mathématique basée sur un modèle à double composante d'écoulement et de transport par une onde cinématique a été développée pour évaluer la distribution dynamique en isotopes dans les précipitations et dans l'écoulement en région aride. Ce modèle décrit le transport et les variations des δ18O de la pluie vers le ruissellement et l'écoulement de surface dans un bassin aride rocheux où le stockage se fait dans des dépressions peu profondes uniformément réparties. Le problème a été résolu numériquement. Le modèle a été calibré au moyen d'une chronique de débits et d'une distribution des δ18O dans la pluie et dans l'écoulement de surface sur un petit bassin versant rocheux du site expérimental de Sede Boker (Israël). La simulation d'un résultat crédible par rapport aux observations a été obtenu après un ajustement des paramètres par une méthode d'essais et d'erreurs. L'analyse de sensibilité et l'application du modèle ont ensuite été réalisés. Le modèle est plutôt sensible aux changements des paramètres caractérisant les zones de stockage dans les dépressions. Le modèle rend compte de l'effet de mémoire isotopique dans l'eau dans le stockage des dépressions entre les événements séquentiels de pluie. L'utilisation d'un modèle à double composante d'écoulement et de transport par onde cinématique permet un ajustement qualitatif et quantitatif adapté entre les distributions des δ18O calculées et observées dans l'écoulement de surface. Resumen Un nuevo método matemático basado en un modelo de doble componente de onda cinemática de flujo y transporte permite caracterizar la distribución isotópica dinámica de las tormentas en zonas áridas y la escorrentía. Este modelo describe el transporte y la evolución del δ18O en la lluvia, flujo superficial y escorrentía en una cuenca rocosa de clima árido con detención superficial distribuida uniformemente. El modelo se calibró numéricamente utilizando un conjunto de datos de descarga temporal y de distribución de δ18O para lluvia y escorrentía recogida en una pequeña cuenca rocosa en el Centro de Experimentación de Sede Boker, Israel. Se obtuvo un buen ajuste a los datos tras un ajuste de parámetros mediante prueba y error. Se realizó un análisis de sensibilidad que indicó que el modelo resulta ser bastante sensible a cambios en los parámetros que caracterizan las zonas de baja detención superficial. El modelo también refleja el efecto de la memoria isotópica en el agua de estas zonas de detención entre los distintos periodos de lluvias. El uso de un modelo de doble componente de onda cinemática de flujo y transporte proporciona un buen ajuste cualitativo y cuantitativo entre los datos medidos y calculados de δ18O en la escorrentía.

National Workshop on Astrobiology: The Life Science Involvement of AAS I Laben

NASA Astrophysics Data System (ADS)

Adami, Giorgio

2006-12-01

The search for traces of past and present life is a complex and multidisciplinary research activity involving several scientific heritages and a specific industrial ability for planetary exploration. Laben was established in 1958 to design and manufacture electronic instruments for research in nuclear physics. In the mid 2004 the company was merged with Alenia Spazio. It is now part of Alcatel Alenia Space, a French Italian joint venture. Alcatel Alenia Space Italia SpA is a Finmeccanica Company. Currently the plant of Vimodrone provides a wide heritage in life science oriented to space application. The experience in Space Life Science is consolidated in the following research areas: (1) Physiology: Mouse models related to studies on human physiology Human neuroscience research and dosimetry (2) Animal Adaptation and Behaviour: mice behaviour related to stabling stress (3) Developmental Biology: aquatic microorganisms cultivation (4) Cell culture & Biotechnology: Protein crystal growth General purpose Multiwell Next Biotechnology studies and development: Bio reactor, mainly oriented to tissue engineering Microsensor for tissue control (organ replacement) Multiwell for adherent cell culture or for automated biosensor based on cell culture Experiment Container for organic systems Experiment Container for small animals Instrumentation based on fluorescent Biosensors Sensors for Life science experiments for Biopan capsule and Space Vehicle Ray Shielding Materials Random Positioning Machine specialisation (Support ground equipment) The biological features of this heritage is at disposal for the exobiology multi science. The involvement of industries, from the beginning of the exobiology projects, allows a cost effective technologies closed loop development between Research Centres, Principal Investigators and industry.

Confirmatory factor analysis of the Appraisal of Self-Care Agency Scale - Revised.

PubMed

Stacciarini, Thaís Santos Guerra; Pace, Ana Emilia

2017-01-30

to analyze the factor structure of the Appraisal of Self-Care Agency Scale-Revised (ASAS-R), adapted for Brazil. methodological study conducted with 150 individuals with diabetes mellitus cared for by the Family Health Strategy, most of whom are elderly with low educational levels. The test of the hypothesis concerning the confirmatory factor composition of the ASAS-R was performed using latent variables structural equations. the model's goodness-of-fit indexes were satisfactory (χ2 = 259.19; χ2/g.l = 2.97, p < 0.001; GFI = 0.85; RMR = 0.07; RMSEA = 0.09); the factor loads were greater than 0.40; and most item-to-factor-correlations presented moderate to strong magnitude (0.34 to 0.58); total alpha value was 0.74, while the alpha of the three factors were 0.69, 0.38 and 0.69, respectively. the scale's factor structure presented satisfactory validity and reliability results, with the exception of one factor. Application of this scale to samples of the general population is desirable in order to strengthen analyses of internal consistency and the dimensionality of the factor structure. This study is expected to contribute to further studies addressing the self-care agency construct and the development of the ASAS-R. analisar a estrutura fatorial da escala de avaliação da capacidade de autocuidado, Appraisal of Self Care Agency Scale-Revised (ASAS-R), adaptada no Brasil. estudo metodológico conduzido em 150 usuários com diabetes mellitus, maioria idosos e com baixa escolaridade, em seguimento na Estratégia Saúde da Família. O teste de hipótese da composição fatorial confirmatória da escala ASAS-R foi realizado via modelo de equações estruturais para variáveis latentes. os valores dos índices de ajuste do modelo foram satisfatórios (χ2 de 259,19; χ2/g.l de 2,97, p < 0,001; GFI = 0,85; RMR = 0,07; RMSEA = 0,09), as cargas fatoriais foram superiores a 0,40, maioria das correlações item e fator foi de moderada a forte magnitude (0,34 a 0,58) e os valores de alfa total de 0,74 e dos três fatores de 0,69, 0,38 e 0,69, respectivamente. estrutura fatorial da escala com resultados satisfatórios de validade e de confiabilidade, exceto um de seus fatores. É desejável que essa Escala seja aplicada em amostras da população geral, para fortalecer as análises de consistência interna e de dimensionalidade da estrutura fatorial, e espera-se que este estudo possa contribuir no avanço de outras pesquisas que trabalham com o construto de capacidade de autocuidado e no desenvolvimento da Escala ASAS-R. analizar la estructura factorial de la escala de evaluación de la capacidad de autocuidado, Appraisal of Self Care Agency Scale-Revised (ASAS-R), adaptada en Brasil. estudio metodológico conducido en 150 usuarios con diabetes mellitus, mayoría de ancianos y con baja escolaridad, acompañados por la Estrategia Salud de la Familia. El test de hipótesis de la composición factorial confirmatoria de la escala ASAS-R fue realizado por medio del modelo de ecuaciones estructurales para variables latentes. los valores de los índices de ajuste del modelo fueron satisfactorios (χ2 de 259,19; χ2/g.l de 2,97, p < 0,001; GFI = 0,85; RMR = 0,07; RMSEA = 0,09), las cargas factoriales fueron superiores a 0,40, la mayoría de las correlaciones ítem y factor varió de moderada a fuerte magnitud (0,34 la 0,58); valores de alfa total de 0,74 y de los tres factores de 0,69, 0,38 y 0,69, respectivamente. la estructura factorial de la escala obtuvo resultados satisfactorios de validez y de confiabilidad, excepto uno de sus factores. Es deseable que esa escala sea aplicada en muestras de la población general, para fortalecer los análisis de consistencia interna y de dimensionalidad de la estructura factorial; y se espera que este estudio pueda contribuir para avanzar en nuevas investigaciones que vengan a trabajar con el constructo de capacidad de autocuidado y con el desarrollo de la escala ASAS-R.

Scale of attitudes toward alcohol - Spanish version: evidences of validity and reliability.

PubMed

Ramírez, Erika Gisseth León; Vargas, Divane de

2017-08-03

validate the Scale of attitudes toward alcohol, alcoholism and individuals with alcohol use disorders in its Spanish version. methodological study, involving 300 Colombian nurses. Adopting the classical theory, confirmatory factor analysis was applied without prior examination, based on the strong historical evidence of the factorial structure of the original scale to determine the construct validity of this Spanish version. To assess the reliability, Cronbach's Alpha and Mc Donalid's Omega coefficients were used. the confirmatory factor analysis indicated the good fit of the scale model in a four-factor distribution, with a cut-off point at 3.2, demonstrating 66.7% of sensitivity. the Scale of attitudes toward alcohol, alcoholism and individuals with alcohol use disorders in Spanish presented robust psychometric qualities, affirming that the instrument possesses a solid factorial structure and reliability and is capable of precisely measuring the nurses' atittudes towards the phenomenon proposed. validar a Escala de atitudes frente ao álcool, ao alcoolismo e a pessoas com transtornos relacionados ao uso do álcool, versão espanhola. estudo metodológico, realizado com 303 enfermeiros colombianos. Seguindo a teoria clássica, foi aplicada a análise fatorial confirmatória sem exploração preliminar, com base na forte evidência histórica da estrutura fatorial do instrumento original para a validação de construto desta versão em espanhol. Para a avaliação da confiabilidade foram utilizados os coeficientes de Alfa de Cronbach e Ômega de Mc Donald. a análise fatorial confirmatória indicou o bom ajuste do modelo da escala na distribuição de quatro fatores, compreendendo 48 itens em sua versão espanhola. Os índices de confiabilidade foram satisfatórios, com ponto de corte observado em 3,2, demonstrando sensibilidade de 66,7%. a Escala de atitudes frente ao álcool, ao alcoolismo e a pessoas com transtornos relacionados ao uso do álcool no idioma espanhol, apresentou qualidades psicométricas robustas, afirmando que se trata de um instrumento com estrutura fatorial e confiabilidade sólidas, capaz de medir com precisão as atitudes dos enfermeiros frente ao fenômeno proposto. validar la Escala de actitudes frente al alcohol, al alcoholismo y a la persona con trastornos relacionados al uso de alcohol en su versión española. estudio de tipo metodológico, realizado con 303 enfermeros colombianos. Siguiendo la teoría clásica, se utilizó análisis factorial confirmatoria sin exploración previa, con base en la fuerte evidencia histórica de la estructura factorial del instrumento original para determinar la validez de constructo de esta versión en español. La confiabilidad se evaluó por los coeficientes de Alfa de Cronbach y Omega de Mc Donald. el análisis factorial confirmatorio indicó buen ajuste del modelo de la escala en la distribución de cuatro factores, comportando 48 ítems en su versión en español. Los índices de confiabilidad fueron satisfactorios, con un punto de corte observado en 3,2, mostrando sensibilidad del 66,7%. la Escala de actitudes frente al alcohol, al alcoholismo y a la persona con trastornos relacionados al uso de alcohol en idioma español presentó cualidades psicométricas robustas, afirmando que se trata de un instrumento con estructura factorial y confiabilidad sólidas, capaz de medir con precisión las actitudes de los enfermeros frente al fenómeno propuesto.

E-mail et Web: Pour une Navigation Sans Risque

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dellabella, Sebastien

2010-06-28

Présentation orale en français, support visuel en anglais. À travers des exemples concrets, vous consoliderez vos connaissances et pourrez ainsi réajuster vos habitudes concernant l’utilisation sécurisée de votre boîte e-mail et de votre navigateur Web.

Psychometric Testing of INTEGRARE, an Instrument for the Assesment of Pressure Ulcer Risk in Inpatients.

PubMed

Porcel-Gálvez, Ana María; Romero-Castillo, Rocío; Fernández-García, Elena; Barrientos-Trigo, Sergio

2017-08-21

The aim of this study is to evaluate the psychometric properties of INTEGRARE, an instrument based on Nursing Outcome Classification. A multicenter, cross-sectional, methodological design was used. The study included 3,835 patients. Internal consistency α = 0.86. Confirmatory factor analysis demonstrated the unidimensionality of the scale, indicating a good model fit (CMIN/DF = 4; GFI, CFI, NFI, IFI = 0.999; RMSEA = 0.028). INTEGRARE is a valid and reliable instrument with high sensitivity, specificity, and diagnostic accuracy in measuring pressure ulcer (PU) risk in inpatients. This instrument allows us to know the effectiveness of nursing interventions, providing evidence for the validation of the diagnosis Risk for pressure ulcer (00249) as well as on health outcomes, due to the fact that PUs are nursing-sensitive outcomes. Evaluar las propiedades psicométricas de INTEGRARE, un instrumento basado en la Clasificación de Resultados de Enfermería. MÉTODO: Se optó por un diseño transversal multicéntrico. El estudio incluyó a 3,835 pacientes. Consistencia interna α = 0.86. El análisis factorial confirmatorio demostró la unidimensionalidad de la escala, indicando un buen ajuste del modelo (CMIN/DF = 4; GFI, CFI, NFI, IFI = 0.999; RMSEA = 0.028). INTEGRARE es un instrumento válido y fiable con alta sensibilidad, especificidad y precisión diagnóstica en la medición de riesgo de úlcera por presión (UPP) en pacientes hospitalizados. IMPLICACIONES PARA LA PRÁCTICA ENFERMERA: Este instrumento nos permite conocer la efectividad de las intervenciones enfermeras, aportando evidencia para la validación del diagnóstico Riesgo de úlcera por presión (00249), así como sobre los resultados de salud, debido a que las UPP son resultados sensibles a la práctica enfermera. © 2017 NANDA International, Inc.

Serious game development as a strategy for health promotion and tackling childhood obesity.

PubMed

Dias, Jéssica David; Mekaro, Marcelo Shinyu; Cheng Lu, Jennifer Kaon; Otsuka, Joice Lee; Fonseca, Luciana Mara Monti; Zem-Mascarenhas, Silvia Helena

2016-08-15

to develop and assess a serious game on healthy eating and physical activity to promote health and tackle childhood obesity. a descriptive, applied and methodological study.For the development of the game, the following steps were taken: conceptualization, pre-production with the development of the game documentation, prototyping, production and assessment of thecomputer and health experts. a prototype has been developed up to beta version. The game was positively assessed both in terms of gameplay and mechanics, and in relation to the content presented, standing out as a powerful strategy for health promotion. The information from the assessment phase contributed to the settings in the software in order to make it available in the future for the target population of this research. The greatest advantage of the proposed game is the fact that it is an open educational resource. the expert assessments showed that the game has great educational potential and it is considered suitable for future application to the target audience.The serious game can become a technological teaching resource available for use in schools and health facilities, and can also be reused for the production of other educational games by accessing its source code. desenvolver e avaliar um serious game (Jogo sério) sobre alimentação saudável e exercício físico para promoção da saúde e auxílio ao enfrentamento da obesidade infantil. estudo descritivo, aplicado e metodológico. Para o desenvolvimento do jogo, foram percorridas as seguintes etapas: conceituação, pré-produção com desenvolvimento da documentação do jogo, prototipagem, produção e avaliação de especialistas de computação e saúde. desenvolveu-se um protótipo até a versão beta. O jogo foi avaliado positivamente tanto em relação à jogabilidade e mecânica, quanto em relação ao conteúdo apresentado, destacando-se como uma estratégia potente para a promoção de saúde. As informações oriundas da fase de avaliação serviram de subsídio para adequações no software com a finalidade de disponibilizá-lo futuramente à população alvo da pesquisa. O grande diferencial do jogo proposto é o fato do mesmo ser um recurso educacional aberto. as avaliações com especialistas demonstraram que o jogo tem grande potencial educacional e considerado adequado para a aplicação futura com o público-alvo. O serious game pode ser um recurso didático tecnológico acessível para uso em escolas e unidades de saúde, além de poder ser reutilizado para a produção de outros jogos educacionais através do acesso de seu código fonte. desarrollar y evaluar un serious game (Juego serio) sobre alimentación saludable y ejercicio físico para la promoción de la salud y ayudar en el enfrentamiento de la obesidad infantil. estudio descriptivo, aplicado y metodológico. Para el desarrollo del juego, se iniciaron los siguientes pasos: conceptualización, pre-producción con el desarrollo de la documentación del juego, creación de prototipos, producción y evaluación por expertos informáticos y en salud. un prototipo se ha desarrollado hasta la versión beta. El juego se evaluó positivamente tanto en términos de jugabilidad y mecánica como en relación con el contenido presentado, destacándose como una poderosa estrategia para la promoción de la salud. La información de la fase de evaluación contribuyó a los ajustes en el software con el fin de que esté disponible en el futuro para la población objetivo de esa investigación. La gran vantaja del juego propuesto es el hecho de que es un recurso educativo abierto. las evaluaciones de los expertos mostraron que el juego tiene un gran potencial educativo y es considerado adecuado para la aplicación futura para el público objetivo. El serious game puede convertirse en un recurso para la enseñanza tecnológica disponible para uso en las escuelas y en los centros de salud, y también puede ser reutilizado para la producción de otros juegos educativos mediante el acceso a su código fuente.

Modelo analítico del efecto de PRS sobre satélites GPS

NASA Astrophysics Data System (ADS)

Meza, A.; Brunini, C.; Usandivaras, J. C.

El sistema GPS (Global Position System) es, hoy en día, la herramienta de navegación y posicionamiento más potente y lo será sin duda en la próxima década. Gran parte de su valiosa utilidad se debe a la alta precisión que permite lograr y ésta, a su vez, depende, entre otras causas, de la precisión con que se conocen las órbitas de los satélites. La presión de radiación solar (PRS) fija el límite de la precisión con que pueden calcularse en la actualidad las efemérides satelitarias. El objetivo de este trabajo es proponer una mejor resolución de este fenómeno. El modelo analítico aquí presentado, se basa en el análisis del comportamiento de los residuos de un ajuste por mínimos cuadrados en el que se utiliza el modelo de PRS propuesto por Beutler. El mismo consiste en un modelo determinista del fenómeno con dos parámetros libres. Los resultados obtenidos ponen de manifiesto que, aún después de aplicar dichos parámetros, prevalecen en los residuos efectos semidiurnos en las componentes radial,tangencial y normal. Estos resultados obtenidos se comparan con los de un trabajo desarrollado por el Instituto de Berne (Beutler et al., 1994), en el que se utilizaron como pseudo-observaciones las órbitas precisas del IGS (CODE). El intervalo de integración escogido por este centro fueron las semanas 680 y 681. En resumen se tienen arcos de 14 días para todos los satélites, donde las efemérides precisas de los mismos para los 14 días fueron utilizados como pseudo-observaciones. El modelo de fuerza que empleó dicho centro fue básicamente el tradicional en lo que respecta al modelo de las fuerzas gravitacionales, y para la PRS utilizo el modelo standard de Beutler. Los parámetros de este modelo junto con las 6 condiciones iniciales (posición y velocidad) fueron ajustados por el método general de mínimos cuadrados. Los residuos en la componente radial, tangencial y normal, para los satélites con un buen comportamiento, presentan una componente semidiurna. El modelo analítico planteado en este trabajo, predice el comportamiento de los residuos que se observan en las publicaciones más recientes. Esto abre el camino para plantear una estimación distinta de las incógnitas del problema, basado en el método de colocación por mínimos cuadrados. Ello requiere modelar estadísticamente la señal debida a las componentes de la PRS que no son tomadas en cuenta en el modelo determinista.

Patient's breath controls comfort devices

NASA Technical Reports Server (NTRS)

Schrader, M.; Carpenter, B.; Nichols, C. D.

1972-01-01

Patient assist system for totally disabled persons was developed which permits a person, so paralyzed as to be unable to move, to activate by breathing, a call system to summon assistance, turn the page of a book, ajust his bed, or do any one of a number of other things. System consists of patient assist control and breath actuated switch.

E-mail et Web: Pour une Navigation Sans Risque

ScienceCinema

Dellabella, Sebastien

2018-04-27

Présentation orale en français, support visuel en anglais. À travers des exemples concrets, vous consoliderez vos connaissances et pourrez ainsi réajuster vos habitudes concernant l’utilisation sécurisée de votre boîte e-mail et de votre navigateur Web.

Profile of the newly graduated physicians in southern Brazil and their professional insertion.

PubMed

Purim, Kátia Sheylla Malta; Borges, Luiza DE Martino Cruvinel; Possebom, Ana Carolina

2016-01-01

Knowledge of the profile and professional integration of new graduates enables adjustments in medical education. This study evaluated 107 graduates from a private institution in the Brazilian South region, using a self-administered electronic questionnaire. There were similar participation of young physicians of both genders and higher male concentration in general surgery. Graduates are inserted in the public and private labor market. Most do extra shifts in emergency services and trauma surgery, where there is greater need for clinical and surgical skills. These findings suggest that adequate surgical training during graduation is critical to employability. RESUMO O conhecimento do perfil e inserção profissional dos recém-formados possibilita ajustes na educação médica. Este estudo avaliou 107 egressos de instituição privada no sul do país, utilizando questionário eletrônico autoaplicável. Houve participação similar de jovens de ambos os sexos e maior concentração masculina na área de cirurgia geral. Os egressos estão inseridos no mercado de trabalho público e privado. A maioria faz plantões extras em serviços de emergência e cirurgia do trauma, onde há maior necessidade de habilidades clinicas e cirúrgica. Esses achados apontam que a formação cirúrgica adequada durante a graduação é fundamental para a empregabilidade.

Methodological proposal for validation of the disinfecting efficacy of an automated flexible endoscope reprocessor.

PubMed

Graziano, Kazuko Uchikawa; Pereira, Marta Elisa Auler; Koda, Elaine

2016-08-08

to elaborate and apply a method to assess the efficacy of automated flexible endoscope reprocessors at a time when there is not an official method or trained laboratories to comply with the requirements described in specific standards for this type of health product in Brazil. the present methodological study was developed based on the following theoretical references: International Organization for Standardization (ISO) standard ISO 15883-4/2008 and Brazilian Health Surveillance Agency (Agência Nacional de Vigilância Sanitária - ANVISA) Collegiate Board Resolution (Resolução de Diretoria Colegiada - RDC) no. 35/2010 and 15/2012. The proposed method was applied to a commercially available device using a high-level 0.2% peracetic acid-based disinfectant. the proposed method of assessment was found to be robust when the recommendations made in the relevant legislation were incorporated with some adjustments to ensure their feasibility. Application of the proposed method provided evidence of the efficacy of the tested equipment for the high-level disinfection of endoscopes. the proposed method may serve as a reference for the assessment of flexible endoscope reprocessors, thereby providing solid ground for the purchase of this category of health products. propor e aplicar um método para a avaliação da eficácia de processadoras automáticas de endoscópios flexíveis, em um momento em que ainda não existe no Brasil um método oficial, nem tampouco laboratórios capacitados que contemplem os requisitos das normas específicas aplicáveis a esse tipo de produto para a saúde. caracterizou-se como pesquisa metodológica e foi desenvolvido com base em três referenciais teóricos: norma técnica International Organization for Standardization (ISO) - ISO 15883-4/2008, Resolução de Diretoria Colegiada (RDC) nº35/2010 e RDC nº15/2012 da Agência Nacional de Vigilância Sanitária (ANVISA). Aplicou-se o método proposto em um equipamento específico, comercialmente disponível, utilizando desinfetante de alto nível à base de ácido peracético 0,2%. o método de avaliação proposto mostrou-se robusto, à medida que as recomendações das legislações pertinentes ao equipamento avaliado foram incorporadas, com algumas adaptações para sua exequibilidade. A aplicação do método proposto permitiu atestar a eficácia do equipamento utilizado na desinfecção de alto nível de endoscópios. o método pode servir de referência para a avaliação de reprocessadoras de endoscópios flexíveis, subsidiando a aquisição dessa categoria de produtos para a saúde. elaborar y aplicar un método para evaluar la eficacia de reprocesadores automatizados de endoscopios flexibles en un momento en el que no hay un método oficial o laboratorios capacitados para cumplir con los requisitos descritos en las normas específicas para este tipo de producto para la salud en Brasil. el presente estudio metodológico fue desarrollado en base a las siguientes referencias teóricas: Organización Internacional de Normalización (International Organization for Standardization - ISO) norma ISO 15883-4/2008 y Agencia Nacional de Vigilancia de la Salud de Brasil (Agência Nacional de Vigilância Sanitária - ANVISA) Resolución del Directorio Colegiado (Resolução de Diretoria Colegiada - RDC) № 35/2010 y 15/2012. El método propuesto se aplicó a un dispositivo comercialmente disponible usando un desinfectante al 0,2% a base de ácido peracético de alto nivel. el método de evaluación propuesto se evaluó como fuerte después de que las recomendaciones formuladas en la legislación pertinente se incorporaron con algunos ajustes para garantizar su factibilidad. La aplicación del método propuesto proporciona evidencia de la eficacia de los equipos de prueba para la desinfección de alto nivel de endoscopios. el método propuesto puede servir de referencia para la evaluación de reprocesadores de endoscopios flexibles, proporcionando de este modo bases sólidas para la compra de esta categoría de productos de salud.

75 FR 51531 - Proposed Collection; Comment Request for Forms 941, 941-PR, 941-SS, 941-X, 941-X(PR), Schedule B...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-20

... 941, 941-PR, 941- SS, 941-X, 941-X(PR), Schedule B (Form 941), Schedule R (Form 941) and Schedule B... Islands, and the U.S. Virgin Islands), 941-X, Adjusted Employer's Quarterly Federal Tax Return or Claim for Refund, 941-X(PR), Ajuste a la Declaracion Federal Trimestral del Patrono o Reclamacion de...

Association of subclinical inflammation, glycated hemoglobin and risk for obstructive sleep apnea syndrome.

PubMed

D'Aurea, Carolina Vicaria Rodrigues; Cerazi, Bruno Gion de Andrade; Laurinavicius, Antonio Gabriele; Janovsky, Carolina Castro Porto Silva; Conceição, Raquel Dilguerian de Oliveira; Santos, Raul D; Bittencourt, Márcio Sommer

2017-01-01

To investigate the inter-relation between high sensitivity C-reactive protein and glycated hemoglobin in prediction of risk of obstructive sleep apnea. We included all individuals participating in a check-up program at the Preventive Medicine Center of Hospital Israelita Albert Einstein in 2014. The Berlin questionnaire for risk of obstructive sleep apnea was used, and the high sensitivity C-reactive protein and glycated hemoglobin levels were evaluated. The sample included 7,115 participants (age 43.4±9.6 years, 24.4% women). The Berlin questionnaire showed changes in 434 (6.1%) individuals. This finding was associated with high sensitivity C-reactive protein and glycated hemoglobin levels (p<0.001). However, only the association between the Berlin questionnaire result and glycated hemoglobin remained significant in the adjusted multivariate analysis, for the traditional risk factors and for an additional model, including high-density lipoprotein cholesterol and triglycerides. The glycated hemoglobin, even below the threshold for diagnosis of diabetes, is independently associated with obstructive sleep apnea syndrome, even after adjustment for obesity and C-reactive protein. These findings suggest a possible pathophysiological link between changes in insulin resistance and obstructive sleep apnea syndrome, independently from obesity or low-grade inflammation. Investigar a inter-relação entre proteína C-reativa de alta sensibilidade e hemoglobina glicada na predição do risco de apneia obstrutiva do sono. Foram incluídos todos os indivíduos participantes do programa de check-up do Centro de Medicina Preventiva Hospital Israelita Albert Einstein em 2014. Foi aplicado o questionário de Berlin sobre risco de apneia do sono, e avaliadas as dosagens de hemoglobina glicada e proteína C-reativa de alta sensibilidade. Foram incluídos 7.115 participantes (idade 43,4±9,6 anos, 24,4% mulheres). A prevalência de alteração no questionário de Berlin foi de 434 (6,1%). A alteração do questionário de Berlin associou-se positivamente aos resultados da proteína C-reativa de alta sensibilidade e da hemoglobina glicada (p<0,001). No entanto, apenas a associação entre o resultado do questionário de Berlin e a hemoglobina glicada permaneceu significativa na análise multivariada ajustada tanto para fatores de risco tradicionais quanto para um modelo adicional, que incluiu também lipoproteína de alta densidade-colesterol (HDL-c) e triglicérides. A hemoglobina glicada, mesmo em valores abaixo do critério diagnóstico para diabetes mellitus, está associada de forma independente ao risco para síndrome da apneia obstrutiva do sono, mesmo após ajuste para obesidade e proteína C-reativa. Estes achados sugerem possível ligação fisiopatológica entre alterações na resistência insulínica e a síndrome da apneia obstrutiva do sono, que independe da obesidade ou inflamação de baixo grau.

Simplificando a luneta com lente de óculos

NASA Astrophysics Data System (ADS)

Canalle, J. B. G.; de Souza, A. C. F.

2003-08-01

A principal ferramenta de trabalho do astrônomo é o telescópio. O manuseio do mesmo é sempre motivo de enorme curiosidade por parte de alunos do ensino fundamental ou médio e até mesmo dos respectivos professores. Visando propiciar o acesso de alunos e professores ou interessados em geral a uma luneta de fácil construção, com materiais alternativos, de fácil localização no comércio, de baixo custo, resistente ao manuseio de alunos, simplificamos a montagem de uma luneta construída com lente de óculos, de 1 ou 2 graus positivos, e monóculo de fotografia, publicado por este autor no CCEF, vol.11(3), 212, 1994. Esta luneta, a qual permite ver as crateras lunares, apresentava como maior dificuldade de construção o tripé e o local de formação da imagem. Substituímos o tripé de madeira por uma simples garrafa PET de 2 litros cheia d'água. No lugar da ocular usamos a lente do monóculo de fotografia (ou visor de fotografia) encaixado dentro de uma bucha de redução curta, de 40 x 32 mm, e esta não mais dentro de uma luva (conexão hidráulica) de 40 mm de diâmetro, mas sim encostada no próprio tubo de PVC móvel (o qual permite o ajuste do foco) de 40 mm de diâmetro e presa a este por outro tubo de 40 mm de diâmetro e 10 cm de comprimento, serrado ao longo do seu comprimento. Com isto podemos ajustar a posição deste tubo de 10 cm também para que uma das suas extremidades coincida com o local de formação da imagem. Desta forma o observador saberá o local exato da formação da imagem, o que não era evidente na montagem anterior e causava uma dificuldade inicial até se descobrir a posição exata em que se deveria colocar o olho. Deste maneira, a montagem inicial que já era simples ficou ainda mais simples, mais barata e mais confortável para o uso. Um exemplar da mesma será exposto durante a XXIX Reunião Anual da Sociedade Astronômica Brasileira para inspeção e uso dos participantes.

Medición de los parámetros cosmológicos q0, ΩM, y ΩΛ, usando supernovas de Tipo Ia distantes

NASA Astrophysics Data System (ADS)

Clocchiatti, A.; High-Z Sne Search Team

Las supernovas de tipo Ia son una herramienta de gran precisión para la medición de distancias de interés cosmológico. Los métodos recientes de calibración de su luminosidad intrínseca, que hacen uso de la forma de las curvas de luz en varios colores y permiten diferenciar entre supernovas distantes, intrínsecamente débiles, u oscurecidas por extinción, reducen la dispersión del método que las asume de magnitud absoluta constante de 0.50 mag a 0.15 mag, e incrementa el valor de la constante de Hubble de ~55 km s-1 Mpc-1, a 65 km s-1 Mpc-1. A partir de la calibración de las supernovas cercanas, con redshifts menores que 0.1, se pueden obtener distancias precisas a supernovas que explotan a alto redshift. Hemos aplicado estos métodos a 16 supernovas con 0.16 < Z < 0.97, encontrando que sus distancias son, en promedio, entre 10% y 13% (dependiendo del método empleado) mayores que las que uno esperaría en un universo con poca masa (Ω = 0.2), sin una constante cosmológica. Todos los métodos de ajuste de curvas de luz, y selección de subgrupos de la muestra de supernovas observadas, favorecen consistentemente modelos del universo que se expanden eternamente y que tienen una constante cosmológica positiva (ΩΛ > 0), y una aceleración de la expansión al presente (q0 < 0). Hay distintas fuentes de posibles errores sistemáticos que merecen ser analizadas, entre ellas: evolución de la metalicidad y estrellas progenitoras, extinción, bias en la elección de la muestra, amplificación por gravitational lensing, y contaminación de la muestra. Ninguno de estos efectos alcanza para reconciliar los datos con ΩΛ = 0, o q0 > 0.

Influência da termoablação com baixa e alta densidade de energia na junção safeno-femoral, utilizando laser endovenoso 1470 nm

PubMed Central

de Araujo, Walter Junior Boim; Erzinger, Fabiano Luiz; Caron, Filipe Carlos; Nejm, Carlos Seme; Timi, Jorge Rufino Ribas

2017-01-01

Resumo Contexto Faz-se importante o conhecimento técnico dos ajustes de potência e de densidade de energia linear endovenosa (linear endovenous energy density, LEED) adequados para atingir o objetivo final da termoablação endovenosa (endovenous laser ablation, EVLA). Objetivos Avaliar a influência de diferentes LEEDs em termos de patência e presença de refluxo, bem como determinar a evolução clínica. Métodos Foram incluídas 60 veias safenas magnas (VSM). Os pacientes foram randomizados em dois grupos: EVLA com baixa potência (7 W e LEED de 20-40 J/cm) e com alta potência (15 W e LEED de 80-100 J/cm). O acompanhamento com eco-Doppler e escore de severidade clínica venoso (VCSS) foi realizado nos intervalos de 3-5 dias, 30 dias, 180 dias e 1 ano após o procedimento. Resultados Dezoito pacientes (29 membros) tratados com 7W de potência e 13 pacientes (23 membros) com 15 W completaram o estudo. Não houve diferença significativa considerando idade, tempo de cirurgia e o uso de analgésicos, lateralidade, gênero e presença de comorbidades. O LEED médio foi de 33,54 J/cm no grupo de 7 W e de 88,66 J/cm no de 15 W. Ambos apresentaram melhora no VCSS, redução significativa dos diâmetros da JSF e ausência de diferença significativa quanto ao aumento do comprimento do coto da VSM e de refluxo após o tratamento. Conclusões A utilização de maior densidade de energia mostrou-se mais efetiva em relação à estabilização do comprimento do coto da VSM e do refluxo em 6 meses. Fazem-se necessários estudos com um período de acompanhamento maior para fundamentar essa hipótese. PMID:29930650

Zero Calcium Score as a Filter for Further Testing in Patients Admitted to the Coronary Care Unit with Chest Pain.

PubMed

Correia, Luis Cláudio Lemos; Esteves, Fábio P; Carvalhal, Manuela; Souza, Thiago Menezes Barbosa de; Sá, Nicole de; Correia, Vitor Calixto de Almeida; Alexandre, Felipe Kalil Beirão; Lopes, Fernanda; Ferreira, Felipe; Noya-Rabelo, Márcia

2017-06-12

The accuracy of zero coronary calcium score as a filter in patients with chest pain has been demonstrated at the emergency room and outpatient clinics, populations with low prevalence of coronary artery disease (CAD). To test the gatekeeping role of zero calcium score in patients with chest pain admitted to the coronary care unit (CCU), where the pretest probability of CAD is higher than that of other populations. Patients underwent computed tomography for calcium scoring, and obstructive CAD was defined by a minimum 70% stenosis on invasive angiography. In 146 patients studied, the prevalence of CAD was 41%. A zero calcium score was present in 35% of the patients. The sensitivity and specificity of zero calcium score yielded a negative likelihood ratio of 0.16. After logistic regression adjustment for pretest probability, zero calcium score was independently associated with lower odds of CAD (OR = 0.12, 95%CI = 0.04-0.36), increasing the area under the ROC curve of the clinical model from 0.76 to 0.82 (p = 0.006). Zero calcium score provided a net reclassification improvement of 0.20 (p = 0.0018) over the clinical model when using a pretest probability threshold of 10% for discharging without further testing. In patients with pretest probability < 50%, zero calcium score had a negative predictive value of 95% (95%CI = 83%-99%), with a number needed to test of 2.1 for obtaining one additional discharge. Zero calcium score substantially reduces the pretest probability of obstructive CAD in patients admitted to the CCU with acute chest pain. (Arq Bras Cardiol. 2017; [online].ahead print, PP.0-0). A acurácia do escore de cálcio coronário zero como um filtro nos pacientes com dor torácica aguda tem sido demonstrada na sala de emergência e nos ambulatórios, populações com baixa prevalência de doença arterial coronariana (DAC). Testar o papel do escore de cálcio zero como filtro nos pacientes com dor torácica admitidos numa unidade coronariana intensiva (UCI), na qual a probabilidade pré-teste de DAC é maior do que em outras populações. Pacientes foram submetidos a tomografia computadorizada para quantificar o escore de cálcio, DAC obstrutiva foi definida por uma estenose mínima de 70% na cineangiocoronariografia invasiva. Um escore clínico para estimar a probabilidade pré-teste de DAC obstrutiva foi criado em amostra de 370 pacientes, usado para definir subgrupos na definição de valores preditivos negativos do escore zero. Em 146 pacientes estudados, a prevalência de DAC foi 41% e o escore de cálcio zero foi demonstrado em 35% deles. A sensibilidade e a especificidade para escore de cálcio zero resultaram numa razão de verossimilhança negativa de 0,16. Após ajuste com um escore clínico com a regressão logística para a probabilidade pré-teste, o escore de cálcio zero foi preditor independente associado a baixa probabilidade de DAC (OR = 0,12, IC95% = 0,04-0,36), aumentando a área abaixo da curva ROC do modelo clínico de 0,76 para 0,82 (p = 0,006). Considerando a probabilidade de DAC < 10% como ponto de corte para alta precoce, o escore de cálcio aumentou a proporção de pacientes para alta precoce de 8,2% para 25% (NRI = 0,20; p = 0,0018). O escore de cálcio zero apresentou valor preditivo negativo de 90%. Em pacientes com probabilidade pré-teste < 50%, o valor preditivo negativo foi 95% (IC95% = 83%-99%). O escore de cálcio zero reduz substancialmente a probabilidade pré-teste de DAC obstrutiva em pacientes internados em UCI com dor torácica aguda. (Arq Bras Cardiol. 2017; [online].ahead print, PP.0-0).

Simultaneous biodegradation of three mononitrophenol isomers by a tailor-made microbial consortium immobilized in sequential batch reactors.

PubMed

Fu, H; Zhang, J-J; Xu, Y; Chao, H-J; Zhou, N-Y

2017-03-01

The ortho-nitrophenol (ONP)-utilizing Alcaligenes sp. strain NyZ215, meta-nitrophenol (MNP)-utilizing Cupriavidus necator JMP134 and para-nitrophenol (PNP)-utilizing Pseudomonas sp. strain WBC-3 were assembled as a consortium to degrade three nitrophenol isomers in sequential batch reactors. Pilot test was conducted in flasks to demonstrate that a mixture of three mononitrophenols at 0·5 mol l -1 each could be mineralized by this microbial consortium within 84 h. Interestingly, neither ONP nor MNP was degraded until PNP was almost consumed by strain WBC-3. By immobilizing this consortium into polyurethane cubes, all three mononitrophenols were continuously degraded in lab-scale sequential reactors for six batch cycles over 18 days. Total concentrations of ONP, MMP and PNP that were degraded were 2·8, 1·5 and 2·3 mol l -1 during this time course respectively. Quantitative real-time PCR analysis showed that each member in the microbial consortium was relatively stable during the entire degradation process. This study provides a novel approach to treat polluted water, particularly with a mixture of co-existing isomers. Nitroaromatic compounds are readily spread in the environment and pose great potential toxicity concerns. Here, we report the simultaneous degradation of three isomers of mononitrophenol in a single system by employing a consortium of three bacteria, both in flasks and lab-scale sequential batch reactors. The results demonstrate that simultaneous biodegradation of three mononitrophenol isomers can be achieved by a tailor-made microbial consortium immobilized in sequential batch reactors, providing a pilot study for a novel approach for the bioremediation of mixed pollutants, especially isomers present in wastewater. © 2016 The Society for Applied Microbiology.

Hydrochemical tracers in the middle Rio Grande Basin, USA: 1. Conceptualization of groundwater flow

NASA Astrophysics Data System (ADS)

Plummer, L. Niel; Bexfield, Laura M.; Anderholm, Scott K.; Sanford, Ward E.; Busenberg, Eurybiades

Chemical and isotopic data for groundwater from throughout the Middle Rio Grande Basin, central New Mexico, USA, were used to identify and map groundwater flow from 12 sources of water to the basin, evaluate radiocarbon ages, and refine the conceptual model of the Santa Fe Group aquifer system. Hydrochemical zones, representing groundwater flow over thousands to tens of thousands of years, can be traced over large distances through the primarily siliciclastic aquifer system. The locations of the hydrochemical zones mostly reflect the ``modern'' predevelopment hydraulic-head distribution, but are inconsistent with a trough in predevelopment water levels in the west-central part of the basin, indicating that this trough is a transient rather than a long-term feature of the aquifer system. Radiocarbon ages adjusted for geochemical reactions, mixing, and evapotranspiration/dilution processes in the aquifer system were nearly identical to the unadjusted radiocarbon ages, and ranged from modern to more than 30 ka. Age gradients from piezometer nests ranged from 0.1 to 2 year cm-1 and indicate a recharge rate of about 3 cm year-1 for recharge along the eastern mountain front and infiltration from the Rio Grande near Albuquerque. There has been appreciably less recharge along the eastern mountain front north and south of Albuquerque. Des données sur les éléments chimiques et les isotopes présents dans l'eau souterraine prélevée à divers endroits dans le bassin moyen du Rio Grande, au centre du Nouveau-Mexique (É-U), ont permis de déterminer l'existence et l'étendue de douze sources d'eau régionales dans le bassin, d'évaluer les âges radiocarbones et de raffiner le modèle conceptuel du système aquifère du groupe de Santa Fe. Des zones hydro-chimiques qui représentent l'écoulement de l'eau souterraine depuis des dizaines de milliers d'années peuvent être suivies sur de longues distances à travers l'aquifère principalement siliclastique. La position des zones hydro-chimiques reflète principalement la distribution moderne des charges hydrauliques mais est incohérente avec une dépression dans le niveau d'eau dans la partie centre-ouest du bassin, ce qui indique que cette dépression est un élément transitoire du système aquifère plutôt qu'un élément à long terme. Les âges radiocarbones ajustés aux réactions géochimiques et aux processus de mélange et d'évapotranspiration/dilution qui ont lieu dans l'aquifère sont presque identiques aux âges non ajustés et varient de la période moderne jusqu'à 30 ka. Les gradients d'âge établis à partir des nids de piézomètres s'étendent de 0.1 à 2 a cm-1 et suggèrent un taux de recharge d'environ 3 cm a-1 le long du front des montagnes à l'est et pour l'infiltration provenant du Rio Grande près d'Albuquerque. Il y a eu substantiellement moins de recharge le long du front des montagnes à l'est, au nord et au sud d'Albuquerque. Se utilizaron datos químicos e isotópicos de agua subterránea a lo largo de la cuenca central del río Grande, Nuevo México, EEUU, para identificar y mapear el flujo de agua subterránea de 12 fuentes de agua a la cuenca para evaluar edades por medio de radio carbon y para refinar el modelo conceptual del sistema acuífero del Grupo Santa Fé. Se puede establecer zonas hidrotérmicas que representan el flujo de agua subterránea a lo largo de miles a miles de decenas de años en grandes distancias a través del sistema acuífero principalmente siliclástico. Las ubicaciones de las zonas hidroquímicas mayormente reflejan la distribucion de la cabeza hidráulica pre-desarollo moderna pero son inconsistentes con una depresión en los niveles de agua pre-desarollo en la zona central oeste de la cuenca. Esto indica que esta depresión es un rasgo transitorio y no un rasgo de largo plazo del sistema acuífero. Las edades de radio carbon ajustadas para los procesos de reaciones geoquímicas, de mezclado y de evapotranspiración-dilución son casi idénticas a los edades de radio carbon no ajustadas oscilan en un rango desde la modernidad a 30 mil años. Las gradientes de edad de nidos de piezometros van de 0.1 a 2 años cm-1 e indican un sitio de recarga de aproximadamente 3 cm/yr para la recarga a lo largo del frente montañoso oriental e infiltración del río Grande cerca de Albuquerque. Se aprecia una recarga menor a lo largo del frente oriental de montañas al norte y al sur de Albuquerque.

Common mental disorders and associated factors: a study of women from a rural area.

PubMed

Parreira, Bibiane Dias Miranda; Goulart, Bethania Ferreira; Haas, Vanderlei José; Silva, Sueli Riul da; Monteiro, Juliana Cristina Dos Santos; Gomes-Sponholz, Flávia Azevedo; Parreira, Bibiane Dias Miranda; Goulart, Bethania Ferreira; Haas, Vanderlei José; Silva, Sueli Riul da; Monteiro, Juliana Cristina Dos Santos; Gomes-Sponholz, Flávia Azevedo

2017-05-25

Identifying the prevalence of Common Mental Disorders and analyzing the influence of sociodemographic, economic, behavioral and reproductive health variables on Common Mental Disorders in women of childbearing age living in the rural area of Uberaba-MG, Brazil. An observational and cross-sectional study. Socio-demographic, economic, behavioral and reproductive health instruments were used, along with the Self-Reporting Questionnaire (SRQ-20) to identify common mental disorders. Multiple logistic regression was used for multivariate data analysis. 280 women participated in the study. The prevalence of Common Mental Disorders was 35.7%. In the logistic regression analysis, the variables of living with a partner and education level were associated with Common Mental Disorders, even after adjusting for the other variables. Our findings evidenced an association of social and behavioral factors with Common Mental Disorders among rural women. Identification and individualized care in primary health care are essential for the quality of life of these women. Identificar a prevalência do transtorno mental comum e analisar a influência de variáveis sociodemográficas, econômicas, comportamentais e de saúde reprodutiva sobre o transtorno mental comum em mulheres em idade fértil, residentes na zona rural do município de Uberaba-MG, Brasil. Estudo observacional e transversal. Foram utilizados instrumentos de caracterização sociodemográfica, econômica, comportamental e de saúde reprodutiva, e o Self-Reporting Questionnaire (SRQ-20) para identificar os transtornos mentais comuns. Na análise multivariada dos dados, foi utilizada a regressão logística múltipla. Participaram do estudo 280 mulheres. A prevalência do transtorno mental comum foi de 35,7%. Na análise de regressão logística, as variáveis convivência com o companheiro e escolaridade, associaram-se ao transtorno mental comum, mesmo após o ajuste para as demais variáveis. Os achados evidenciaram a associação de fatores sociais e comportamentais com o transtorno mental comum, entre mulheres rurais. A identificação e a assistência individualizada na atenção primária de saúde são essenciais para a qualidade de vida destas mulheres. Identificar la prevalencia de trastornos mentales comunes y analizar la influencia de las variables socio-demográficas, económicas, de comportamiento y de salud reproductiva en el trastorno mental común en las mujeres en edad fértil, que viven en el municipio rural de Uberaba, Minas Gerais, Brasil. Estudio observacional y transversal. Se usaron instrumentos sociodemográficos, económicos, de comportamiento y salud reproductiva, y el Self-Reporting Questionnaire (SRQ-20) para identificar los trastornos mentales comunes. En el análisis multifactorial de los datos, se utilizó la regresión logística múltiple. El estudio incluyó a 280 mujeres. La prevalencia de los trastornos mentales comunes fue de 35,7%. En el análisis de regresión logística, las variables convivencia con su pareja y la escolarización se asociaron con trastorno mental común, incluso después de ajustar por otras variables. Los resultados muestran la relación entre los factores sociales y de comportamiento con el trastorno mental común entre las mujeres rurales. La identificación y la atención individual en la atención primaria de salud son esenciales para la calidad de vida de las mujeres.

PSR J0751+1807: un ajuste a los parámetros característicos del sistema binario

NASA Astrophysics Data System (ADS)

De Vito, M. A.; Benvenuto, O. G.

PSR J0751+1807 is a millisecond pulsar belonging to a binary system with a low mass white dwarf companion. This system belongs to the group of recycled pulsars by mass transfer from a close companion, accelerating the pulsar rotation in this process. The orbital period for the system is of 6 hours. In this work we show our fit to the characteristic parameters of the system presented by Nice et al. (2005) FULL TEXT IN SPANISH

Combustion of Solid Propellants (La Combustion des Propergols Solides)

DTIC Science & Technology

1991-07-01

cin~tiques initiales. Il relatives A Ia granulom ~trie ct la surface eat s~me possible dWaller plus loin et sp~cifique des catalyseurs existent, il est...grand nombro do vari~t~s granulom ~triques des proporgols. On pout ainsi observer uno mont donc utilis~es induatriellemont pour notte influence du temps...et do la ajuster la vitosso des vari~tds do tempdrature do laminage sur la diminution granulom ~trie moyenno 400, 200, 100, 10, 3 de l’exposant do

HOW TO ACHIEVE AND MAINTAIN NOTE 6: POSTGRADUATE PROGRAM IN TRANSLATIONAL SURGERY - UNIFESP.

PubMed

Sabino-Neto, Miguel; Ferreira, Lydia Masako

2015-01-01

To show the way to reach and stay in note 6 in the evaluation process of Medicine III of CAPES. Capes determinations were reviewed concerning this topic, grades 6 and 7, and also the difficulties and facilities of running a program that amounted to Note 6 after restructuring and being in compliance with regulations. The main points to achieve and maintain Note 6 were: 1) regular production of master's and doctoral theses with appropriate distribution among all teachers; 2) average time of appropriate titration, as well as strict selection of students who resets the withdrawals and cancellations; 3) production of scientific articles in high impact journals and with academic and student participation in most part; 4) progressive and substantial increase in fundraising and patent search; 5) progressive increase in international exchanges with joint production; 6) visibility through new bilingual website and updated weekly; 7) numerous solidarity activities in research, but also in health services for the population and even in basic education; 8) rigorous selection of students (through design analysis, curriculum and teacher training program); 9) maintenance of high levels teachers production; 10) preparing new teachers for guidance through participation as co-supervision and involvement in the program to fit the needs. The Postgraduate Program in Translational Surgery went through difficult times; was submitted to a series of measures, adjustments, cooperation and understanding of the teaching staff, that took the program from note 3 - and almost closing - to a level of excellence keeping note 6 for three consecutive three-year periods of evaluation. Mostrar o caminho para alcançar e se manter na nota 6 no processo de avaliação da Medicina III da Capes. Foram revisadas as determinações da Capes concernentes ao tema, conceitos 6 e 7, e também as dificuldades e facilidades próprias da execução de um programa que ascendeu à nota 6 após reestruturação e adequação às normas vigentes. Os pontos principais para alcançar e manter conceito 6 foram: : 1) produção regular de teses de mestrado e doutorado com distribuição adequada entre todos os orientadores; 2) tempo médio de titulação adequado, assim como rígida seleção do alunado que zera as desistências e cancelamentos; 3) produção de artigos científicos alta em revistas de impacto e com participação docente e discente na sua grande maioria; 4) aumento progressivo e substancial da captação de recursos e busca de patentes; 5) aumento progressivo dos intercâmbios internacionais com produção conjunta; 6) visibilidade através de novo site bilíngue e atualizado semanalmente; 7) inúmeras atividades de solidariedade em pesquisa, mas também em Serviços de Saúde para a população e mesmo no ensino básico; 8) rigorosa seleção dos alunos (através de análise de projeto, currículo e programa de estágio docente); 9) manutenção de índices elevados de produção dos docentes; 10) preparação de novos docentes para orientação através de participação como co-orientação e envolvimento no programa para se adequar às necessidades. O Programa de Pós-Graduação em Cirurgia Translacional passou por momentos difíceis e que através de uma série de medidas, ajustes e a colaboração e compreensão do seu corpo docente pode sair da nota 3 e quase fechamento, para um nível de excelência mantendo nota 6 em três triênios seguidos.

Compact light-emitting diode optical fiber immobilized TiO2 reactor for photocatalytic water treatment.

PubMed

O'Neal Tugaoen, Heather; Garcia-Segura, Sergi; Hristovski, Kiril; Westerhoff, Paul

2018-02-01

A key barrier to implementing photocatalysis is delivering light to photocatalysts that are in contact with aqueous pollutants. Slurry photocatalyst systems suffer from poor light penetration and require post-treatment to separate the catalyst. The alternative is to deposit photocatalysts on fixed films and deliver light onto the surface or the backside of the attached catalysts. In this study, TiO 2 -coated quartz optical fibers were coupled to light emitting diodes (OF/LED) to improve in situ light delivery. Design factors and mechanisms studied for OF/LEDs in a flow-through reactor included: (i) the influence of number of LED sources coupled to fibers and (ii) the use of multiple optical fibers bundled to a single LED. The light delivery mechanism from the optical fibers into the TiO 2 coatings is thoroughly discussed. To demonstrate influence of design variables, experiments were conducted in the reactor using the chlorinated pollutant para-chlorobenzoic acid (pCBA). From the degradation kinetics of pCBA, the quantum efficiencies (Φ) of oxidation and electrical energies per order (E EO ) were determined. The use of TiO 2 coated optical fiber bundles reduced the energy requirements to deliver photons and increased available surface area, which improved Φ and enhanced oxidative pollutant removal performance (E EO ). Copyright © 2017 Elsevier B.V. All rights reserved.

Estudio epidemiológico de sucesos traumáticos, trastorno de estrés post-traumático y otros trastornos psiquiátricos en una muestra representativa de Chile

PubMed Central

Pérez Benítez, Carlos I.; Vicente, Benjamin; Zlotnick, Caron; Kohn, Robert; Johnson, Jennifer; Valdivia, Sandra; Rioseco, Pedro

2010-01-01

RESUMEN Durante la década de 1990 en los Estados Unidos (EU), el conocimiento sobre el trastorno de estrés post-traumático (TEPT) evolucionó de estudios específicos en un principio, sobre veteranos de guerra y sobre víctimas de desastres, a estudios epidemiológicos más tarde, sin embargo, la epidemiología del TEPT en países en desarrollo ha sido un área poco estudiada hasta ahora. Los expertos en el área de trauma han propuesto que los sucesos traumáticos que ocurren en la niñez son más perjudiciales para la salud mental que aquellos que ocurren más tarde en la vida. Este trabajo revisa los resultados de un estudio epidemiológico llevado a cabo en Chile. Específicamente, se revisan los resultados sobre las tasas de prevalencia del TEPT, traumas asociados más frecuentemente con él, así como la comorbilidad de este trastorno con otros trastornos psiquiátricos a lo largo de la vida. Igualmente se analizaron las diferencias del TEPT en cada sexo, así como la exposición a traumas en una muestra representativa de chilenos. Además se comparó la prevalencia de trastornos psiquiátricos en personas que sufrieron su primer trauma durante la niñez, durante la edad adulta, o que no reportaron traumas durante su vida. En estos estudios epidemiológicos se usaron módulos del TEPT y trastorno de personalidad antisocial (TPA) de la entrevista diagnóstica siguiendo los criterios del DSM-III-R (DIS–III-R). Para evaluar el resto de los trastornos psiquiátricos se usó la Entrevista Diagnóstica Internacional Compuesta (CIDI). Estos instrumentos fueron administrados en tres ciudades chilenas a 2390 personas mayores de 15 años. Para estimar los errores estándares (EE) debido al diseño de la muestra y a la necesidad de ajuste se usó el método Taylor de linearización seriada. También se usó un análisis de regresión logística para examinar la relación entre el TEPT, los factores demográficos de riesgo y el tipo de trauma. Además se utilizó la regresión logística multivariada para evaluar si la relación entre el TEPT y el sexo pudiera ser explicada por medio de otros factores de riesgo, así como para calcular las tasas y la oportunidad relativa (razón de productos cruzados) de trastornos psiquiátricos a lo largo de la vida. El primer análisis arrojó que la prevalencia de TEPT a lo largo de la vida fue de 4.4% (2.5% para hombres y 6.2% para mujeres). De los hechos traumáticos reportados, la violación sexual tuvo una correlación más alta con el TEPT que los demás hechos traumáticos. Las mujeres tuvieron más probabilidades de experimentar TEPT que los hombres, después de controlar la variable asalto violento. El segundo análisis evidenció que los que habían sufrido traumas a lo largo de la vida tuvieron mayor probabilidad de tener un trastorno psiquiátrico en comparación con aquellos que no reportaron traumas. También se encontró que los que sufrieron su primer trauma durante la infancia tuvieron más probabilidad de desarrollar trastornos de pánico a lo largo de la vida que aquellos que sufrieron su primer trauma en la edad adulta, independientemente del número de traumas que sufrieron y de las diferencias demográficas. Aunque Chile tiene un contexto histórico-cultural y una economía diferente a otros países en los que se ha estudiado anteriormente la epidemiología del TEPT, el presente estudio reflejó tendencias similares a las reportadas en estudios previos. Los hallazgos expuestos enfatizan la importancia de investigar la prevalencia del TEPT, los patrones de comorbilidad del TEPT y las diferencias de sexo en la prevalencia del TEPT en diferentes países. También estos resultados sugieren que los sucesos traumáticos en la infancia (y no en la adultez) pueden estar relacionados con la ocurrencia de trastornos psiquiátricos específicos. PMID:21113425

Development of a moderator system for the High Brilliance Neutron Source project

NASA Astrophysics Data System (ADS)

Dabruck, J. P.; Cronert, T.; Rücker, U.; Bessler, Y.; Klaus, M.; Lange, C.; Butzek, M.; Hansen, W.; Nabbi, R.; Brückel, T.

2016-11-01

The project for an accelerator based high brilliance neutron source HBS driven by Forschungszentrum Jülich forsees the use of the nuclear Be(p,n) or Be(d,n) reaction with accelerated particles in the lower MeV energy range. The lower neutron production compared to spallation has to be compensated by improving the neutron extraction process and optimizing the brilliance. Design and optimiziation of the moderator system are conducted with MCNP and will be validated with measurements at the AKR-2 training reactor by means of a prototype assembly where, e.g., the effect of different liquid H2 ortho/para ratios will be investigated and controlled in realtime via online heat capacity measurements.

Syntax Score and Major Adverse Cardiac Events in Patients with Suspected Coronary Artery Disease: Results from a Cohort Study in a University-Affiliated Hospital in Southern Brazil.

PubMed

Fuchs, Felipe C; Ribeiro, Jorge P; Fuchs, Flávio D; Wainstein, Marco V; Bergoli, Luis C; Wainstein, Rodrigo V; Zen, Vanessa; Kerkhoff, Alessandra C; Moreira, Leila B; Fuchs, Sandra C

2016-09-01

The importance of coronary anatomy in predicting cardiovascular events is well known. The use of traditional anatomical scores in routine angiography, however, has not been incorporated to clinical practice. SYNTAX score (SXscore) is a scoring system that estimates the anatomical extent of coronary artery disease (CAD). Its ability to predict outcomes based on a baseline diagnostic angiography has not been tested to date. To evaluate the performance of the SXscore in predicting major adverse cardiac events (MACE) in patients referred for diagnostic angiography. Prospective cohort of 895 patients with suspected CAD referred for elective diagnostic coronary angiography from 2008 to 2011, at a university-affiliated hospital in Brazil. They had their SXscores calculated and were stratified in three categories: no significant CAD (n = 495), SXscoreLOW-INTERMEDIATE: < 23 (n = 346), and SXscoreHIGH: ≥ 23 (n = 54). Primary outcome was a composite of cardiac death, myocardial infarction, and late revascularization. Secondary endpoints were the components of MACE and death from any cause. On average, patients were followed up for 1.8 ± 1.4 years. The primary outcome occurred in 2.2%, 15.3%, and 20.4% in groups with no significant CAD, SXscoreLOW-INTERMEDIATE, and SXscoreHIGH, respectively (p < 0.001). All-cause death was significantly higher in the SXscoreHIGH compared with the 'no significant CAD' group, 16.7% and 3.8% (p < 0.001), respectively. After adjustment for confounding factors, all outcomes remained associated with the SXscore. SXscore independently predicts MACE in patients submitted to diagnostic coronary angiography. Its routine use in this setting could identify patients with worse prognosis. A importância da anatomia coronariana na predição de eventos cardiovasculares é bem conhecida. O uso de escores anatômicos tradicionais na cineangiocoronariografia de rotina, entretanto, não foi incorporado à prática clínica. O SYNTAX escore (SXescore) é um sistema de escore que estima a extensão anatômica da doença arterial coronariana (DAC). Sua capacidade para predizer desfechos com base na cineangiocoronariografia diagnóstica de base ainda não foi testada. Avaliar o desempenho do SXescore para predizer eventos cardíacos adversos maiores (MACE) em pacientes encaminhados para cineangiocoronariografia diagnóstica. Coorte prospectiva de 895 pacientes com suspeita de DAC encaminhados para cineangiocoronariografia diagnóstica eletiva de 2008 a 2011, em hospital universitário no Brasil. Os pacientes tiveram seus SXescores calculados e foram estratificados em três categorias: 'sem DAC significativa' (n = 495); SXescoreBAIXO-INTERMEDIÁRIO: < 23 (n = 346); e SXescoreALTO: ≥ 23 (n = 54). O desfecho primário foi composto de morte cardíaca, infarto do miocárdio e revascularização tardia. Os desfechos secundários foram MACE e morte por todas as causas. Em média, os pacientes foram acompanhados por 1,8 ± 1,4 anos. Desfecho primário ocorreu em 2,2%, 15,3% e 20,4% nos grupos 'sem DAC significativa', SXescoreBAIXO-INTERMEDIÁRIO e SXescoreALTO, respectivamente (p < 0,001). Morte por todas as causas foi significativamente mais frequente no grupo de SXescoreALTO comparado ao grupo 'sem DAC significativa', 16,7% e 3,8% (p < 0,001), respectivamente. Após ajuste para fatores de confusão, todos os desfechos permaneceram associados com o SXescore. O SXescore prediz independentemente MACE em pacientes submetidos a cineangiocoronariografia diagnóstica. Seu uso rotineiro nesse contexto poderia identificar pacientes de pior prognóstico.

Biological capacitance studies of anodes in microbial fuel cells using electrochemical impedance spectroscopy.

PubMed

Lu, Zhihao; Girguis, Peter; Liang, Peng; Shi, Haifeng; Huang, Guangtuan; Cai, Lankun; Zhang, Lehua

2015-07-01

It is known that cell potential increases while anode resistance decreases during the start-up of microbial fuel cells (MFCs). Biological capacitance, defined as the apparent capacitance attributed to biological activity including biofilm production, plays a role in this phenomenon. In this research, electrochemical impedance spectroscopy was employed to study anode capacitance and resistance during the start-up period of MFCs so that the role of biological capacitance was revealed in electricity generation by MFCs. It was observed that the anode capacitance ranged from 3.29 to 120 mF which increased by 16.8% to 18-20 times over 10-12 days. Notably, lowering the temperature and arresting biological activity via fixation by 4% para formaldehyde resulted in the decrease of biological capacitance by 16.9 and 62.6%, indicating a negative correlation between anode capacitance and anode resistance of MFCs. Thus, biological capacitance of anode should play an important role in power generation by MFCs. We suggest that MFCs are not only biological reactors and/or electrochemical cells, but also biological capacitors, extending the vision on mechanism exploration of electron transfer, reactor structure design and electrode materials development of MFCs.

Simultaneous Bioreduction of Multiple Oxidized Contaminants Using a Membrane Biofilm Reactor.

PubMed

Li, Haixiang; Lin, Hua; Xu, Xiaoyin; Jiang, Minmin; Chang, Chein-Chi; Xia, Siqing

2017-02-01

  This study tests a hydrogen-based membrane biofilm reactor (MBfR) to investigate simultaneous bioreduction of selected oxidized contaminants, including nitrate (-N), sulfate (), bromate (), chromate (Cr(VI)) and para-chloronitrobenzene (p-CNB). The experiments demonstrate that MBfR can achieve high performance for contaminants bioreduction to harmless or immobile forms in 240 days, with a maximum reduction fluxes of 0.901 g -N/m2·d, 1.573 g /m2·d, 0.009 g /m2·d, 0.022 g Cr(VI)/m2·d, and 0.043 g p-CNB/m2·d. Increasing H2 pressure and decreasing influent surface loading enhanced removal efficiency of the reactor. Flux analysis indicates that nitrate and sulfate reductions competed more strongly than , Cr(VI) and p-CNB reduction. The average H2 utilization rate, H2 flux, and H2 utilization efficiency of the reactor were 0.026 to 0.052 mg H2/cm3·d, 0.024 to 0.046 mg H2/cm2·d, and 97.5% to 99.3% (nearly 100%). Results show the hydrogen-based MBfR may be suitable for removing multiple oxidized contaminants in drinking water or groundwater.

Simulation of regional-scale groundwater flow in the Azul River basin, Buenos Aires Province, Argentina

NASA Astrophysics Data System (ADS)

Varni, Marcelo R.; Usunoff, Eduardo J.

A three-dimensional modular model (MODFLOW) was used to simulate groundwater flow in the Azul River basin, Buenos Aires Province, Argentina, in order to assess the correctness of the conceptual model of the hydrogeological system. Simulated heads satisfactorily match observed heads in the regional water-table aquifer. Model results indicate that: (1) groundwater recharge is not uniform throughout the region but is best represented by three recharge rates, decreasing downgradient, similar to the distribution of soils and geomorphological characteristics; and (2) evapotranspiration rates are larger than previous estimates, which were made by using the Thornthwaite-Mather method. Evapotranspiration rates estimated by MODFLOW agree with results of independent studies of the region. Model results closely match historical surface-flow records, thereby suggesting that the model description of the aquifer-river relationship is correct. Résumé Un modèle modulaire tridimensionnel (MODFLOW) a été utilisé pour simuler les écoulements souterrains dans le bassin de la rivière Azul (Province de Buenos Aires, Argentine), dans le but d'évaluer la justesse du modèle conceptuel du système hydrogéologique. La piézométrie simulée s'ajuste de façon satisfaisante à celle observée pour l'ensemble de la nappe. Les résultats du modèle indiquent que: (1) la recharge de la nappe n'est pas uniforme sur toute la région, mais qu'elle est mieux approchée par trois valeurs différentes, décroissant vers l'aval-gradient, en suivant la même distribution que les sols et les caractéristiques géomorphologiques et (2) l'évapotranspiration est nettement plus importante que prévu initialement à partir de la méthode de Thornthwaite-Mather. Les valeurs d'évapotranspiration fournies par MODFLOW concordent bien avec les résultats d'autres études portant sur la région. Les résultats du modèle reproduisent convenablement les chroniques de débit des écoulements de surface, suggérant ainsi que la description par le modèle des relations rivière-nappe est correcte. Resumen Se ha utilizado el modelo MODFLOW, del Servicio Geológico de los Estados Unidos, para simular el flujo de agua subterránea en la cuenca del arroyo del Azul, Provincia de Buenos Aires, Argentina, con el objeto de evaluar el modelo hidrogeológico conceptual. Los niveles hidráulicos simulados ajustan satisfactoriamente con los niveles observados. Los resultados de la simulación indican que: (1) la recarga no es uniforme, sino que puede caracterizarse con tres zonas en las que sus valores decrecen en la medida en que decrece la pendiente, que guarda similitud con la distribución de suelos y características geomorfológicas y (2) la evapotranspiración sería mayor que la estimada en estudios previos, en los que se utilizó el método de Thornthwaite-Mather. La evapotranspiración estimada mediante la presente simulación concuerda con resultados de varios estudios independientes en la región. Respecto de la relación acuífero-río, existe un muy buen ajuste entre los aportes del acuífero al río simulados y los valores históricos de caudal base.

Usefulness of the Diagnosis "Decreased Cardiac Output (00029)" in Patients With Chronic Heart Failure.

PubMed

Rojas Sánchez, Lyda Zoraya; Hernández Vargas, Juliana Alexandra; Trujillo Cáceres, Silvia Juliana; Roa Díaz, Zayne Milena; Jurado Arenales, Adriana Milena; Toloza Pérez, Yesith Guillermo

2017-10-01

To determine the clinical and construct validity of the nursing diagnosis "decreased cardiac output" (DCO) in patients with chronic heart failure. Cross-sectional study. A total of 200 people were studied. The defining characteristics with the highest prevalence were as follows: arrhythmia (62.5%) and fatigue (61.5%). Adjustment measures such as infit and outfit were maintained between 0.50 and 1.56 and the total variance explained by the measures was 29.3%. This study determined the clinical validity of the nursing diagnosis DCO. Regarding construct validity, adjustment of the defining characteristics to the Rasch model was observed. This study improves the evidence-based practice of nursing and strengthened the role of the nurse who leads care to this population. Determinar la validez clínica y de constructo del diagnóstico de enfermería "Disminución del Gasto Cardíaco" en pacientes con falla cardíaca crónica. MÉTODOS: Estudio de corte transversal. Un total de 200 pacientes fueron estudiados. Las características definitorias con las mayores prevalencias fueron: arritmia (62.5%) y fatiga (61.5%). Medidas de ajuste como el infit y outfit se mantuvieron entre 0.50 y 1.56. El total de la varianza explicada por las medidas fue de 29.3%. Este estudio determinó la validez clínica del diagnóstico de enfermería "Disminución del Gasto Cardíaco". En cuanto a la validez de constructo, se observó que 19 de las 21 características definitorias se ajustaron al modelo Rasch. IMPLICACIONES PARA LA PRÁCTICA DE ENFERMERÍA: Este estudio mejora la práctica basada en la evidencia de enfermería y fortalece el rol de las enfermeras que lideran el cuidado en esta población. © 2016 NANDA International, Inc.

Calibración del retardo ionosférico en observaciones astrométricas y geodésicas a partir de observaciones GPS

NASA Astrophysics Data System (ADS)

Brunini, C.; Kleusberg, A.; Arias, E. F.; de Biasi, M. S.

Los parámetros astrométricos y geodésicos de precisión se determinan hoy mediante la observación con técnicas espaciales (VLBI, GPS y LSR). Las técnicas VLBI y GPS operan en la banda de microondas y en ella la ionósfera terrestre es dispersiva. Las señales que provienen de las radiofuentes y de los satélites atraviesan la ionósfera, donde el índice de refracción difiere de la unidad en una cantidad que es proporcional a la densidad de electrones libres e inversamente proporcional al cuadrado de la frecuencia de la onda portadora. Actualmente el International GPS Service for Geodynamics (IGS) mantiene operacional una red global integrada por más de 50 estaciones equipadas con receptores GPS de alta performance; las observaciones diarias son accesibles a los usuarios mediante ftp. La posibilidad de utilizar estas observaciones en un monitoreo continuo de la ionósfera fue señalada por diversos autores, razón por la cual en los últimos años se ha invertido un significativo esfuerzo en la producción de mapas ionosféricos regionales y globales. En el presente trabajo se utilizan 28 estaciones cuyas observaciones mapean la mayor parte de la ionósfera global. Los resultados obtenidos demuestran la posibilidad de obtener mapas ionosféricos globales con una resolución de medio día y con una precisión de (1.5 nseg (rms)). Dichos mapas proveen valores medios globales para el intervalo ajustado. Los residuos del ajuste por mínimos cuadrados constituyen una señal a partir de la cual pueden estudiarse mejor las variaciones geográficas de la ionósfera y las componentes estadísticas de su variación temporal.

Field technique of permeability tests in highly fissured limestone strata

NASA Astrophysics Data System (ADS)

Al-Salihi, Adnan; Asaad, Abdulah

2002-05-01

Résumé.L'étude de dispositifs de dénoyage est nécessaire pour l'amélioration de sites avant la construction de certaines structures. L'étude de dispositifs de dénoyage efficaces exige d'estimer la valeur du coefficient de perméabilité in situ. Les relations disponibles pour estimer le coefficient de perméabilité ont été développées sur la base de mesures et de conditions de terrain limitées, et les prédictions varient de plusieurs ordres de grandeur. C'est pourquoi il est nécessaire de réaliser des mesures de perméabilité sur le terrain et de déterminer la relation qui permet le meilleur ajustement de ces mesures avant l'étude du dispositif de dénoyage pour des conditions locales et géologiques spécifiques. Ce papier présente des mesures de perméabilité sur le terrain dans des niveaux calcaires complexes chaotiques et diagénétisés. Il propose également une analyse comparative de plusieurs relations disponibles dans la littérature destinées à prédire le coefficient de perméabilité in situ. L'analyse est faite en conditions permanentes et non permanentes. Les résultats montrent que la valeur du coefficient de perméabilité dépend du niveau de la nappe, qui est affecté par le régime de marées. On montre que l'équation de l'US Navy donne la meilleure corrélation avec les mesures de terrain. Resumen.El diseño de sistemas de desecado es necesario para mejorar las condiciones de un emplazamiento antes de la construcción de determinadas estructuras. El diseño de un sistema eficiente de desecado requiere de la estimación del valor de la permeabilidad in-situ. Las relaciones disponibles para tal fin han sido desarrolladas bajo condiciones y medidas de campo limitadas; sus predicciones varían en algunos órdenes de magnitud. Por tanto, es necesario tomar medidas de permeabilidad en campo y determinar la relación que reproduce mejor dichas medidas como paso previo al diseño de un sistema de desecado en condiciones geológicas y de emplazamiento específicas. Este artículo presenta medidas de permeabilidad en campo para estratos de calcita caóticos y diagenéticos. También ofrece un análisis comparativo de diversas relaciones disponibles en la bibliografía con el fin de predecir el valor de la permeabilidad in-situ. El análisis se ha hecho tanto en régimen permanente como en estacionario. Los resultados demuestran que la permeabilidad depende del nivel freático, el cual está afectado por las mareas. La ecuación de la Marina estadounidense es la que proporciona una mejor correlación con las medidas de campo.

Mesoporous Aluminosilicates as a Host and Reactor for Preparation of Ordered Metal Nanowires

NASA Astrophysics Data System (ADS)

Eliseev, A. A.; Napolskii, K. S.; Kolesnik, I. V.; Kolenko, Yu. V.; Lukashin, A. V.; Gornert, P.; Tretyakov, Yu. D.

The creation of functional nanomaterials with the controlled properties is emerging as a new area of great technological and scientific interest, in particular, it is a key technology for developing novel high-density data storage devices. Today, no other technology can compete with magnetic carriers in information storage density and access rate. However, usually very small (10-1000 nm3) magnetic nanoparticles shows para- or superparamagnetic properties, with very low blocking temperatures and no coercitivity at normal conditions. One possible solution of this problem is preparation of highly anisotropic nanostructures. From the other hand, the use of purely nanocrystalline systems is limited because of their low stability and tendency to form aggregates. These problems could be solved by encapsulation of nanoparticles to a chemically inert matrix. One of the promising matrices for preparation of highly anisotropic magnetic nanoparticles is mesoporous silica or mesoporous aluminosilicates. Mesoporous silica is an amorphous SiO2 with a highly ordered uniform pore structure (the pore diameter can be controllably varied from 2 to 50 nm). This pore system is a perfect reactor for synthesis of nanocomposites due to the limitation of reaction zone by the pore walls. One could expect that size and shape of nanoparticles incorporated into mesoporous silica to be consistent with the dimensions of the porous framework.

Relationship between Resting Heart Rate, Blood Pressure and Pulse Pressure in Adolescents.

PubMed

Christofaro, Diego Giulliano Destro; Casonatto, Juliano; Vanderlei, Luiz Carlos Marques; Cucato, Gabriel Grizzo; Dias, Raphael Mendes Ritti

2017-05-01

High resting heart rate is considered an important factor for increasing mortality chance in adults. However, it remains unclear whether the observed associations would remain after adjustment for confounders in adolescents. To analyze the relationship between resting heart rate, blood pressure and pulse pressure in adolescents of both sexes. A cross-sectional study with 1231 adolescents (716 girls and 515 boys) aged 14-17 years. Heart rate, blood pressure and pulse pressure were evaluated using an oscillometric blood pressure device, validated for this population. Weight and height were measured with an electronic scale and a stadiometer, respectively, and waist circumference with a non-elastic tape. Multivariate analysis using linear regression investigated the relationship between resting heart rate and blood pressure and pulse pressure in boys and girls, controlling for general and abdominal obesity. Higher resting heart rate values were observed in girls (80.1 ± 11.0 beats/min) compared to boys (75.9 ± 12.7 beats/min) (p ≤ 0.001). Resting heart rate was associated with systolic blood pressure in boys (Beta = 0.15 [0.04; 0.26]) and girls (Beta = 0.24 [0.16; 0.33]), with diastolic blood pressure in boys (Beta = 0.50 [0.37; 0.64]) and girls (Beta = 0.41 [0.30; 0.53]), and with pulse pressure in boys (Beta = -0.16 [-0.27; -0.04]). This study demonstrated a relationship between elevated resting heart rate and increased systolic and diastolic blood pressure in both sexes and pulse pressure in boys even after controlling for potential confounders, such as general and abdominal obesity. A frequência cardíaca de repouso é considerada um importante fator de aumento de mortalidade em adultos. Entretanto, ainda é incerto se as associações observadas permanecem após ajuste para fatores de confusão em adolescentes. Analisar a relação entre frequência cardíaca de repouso, pressão arterial e pressão de pulso em adolescentes dos dois sexos. Estudo transversal com 1231 adolescentes (716 meninas e 515 meninos, idade de 14-17 anos). Frequência cardíaca, pressão arterial e pressão de pulso foram avaliadas com esfigmomanômetro oscilométrico validado para essa população. Peso e altura foram medidos com balança eletrônica e estadiômetro, respectivamente, e a circunferência abdominal, com uma fita inextensível. Análise multivariada com regressão linear investigou a relação entre frequência cardíaca de repouso, pressão arterial e pressão de pulso em meninos e meninas, controlando para obesidade geral e abdominal. Valores maiores de frequência cardíaca de repouso foram observados em meninas (80,1 ± 11,0 bpm) em comparação a meninos (75,9 ± 12,7 bpm) (p ≤ 0,001). Frequência cardíaca de repouso associou-se com pressão arterial sistólica em meninos [Beta = 0,15 (0,04; 0,26)] e meninas [Beta = 0,24 (0,16; 0,33)], com pressão arterial diastólica em meninos [Beta = 0,50 (0,37; 0,64)] e meninas [Beta = 0,41 (0,30; 0,53)], e com pressão de pulso apenas em meninos [Beta = -0,16 (-0,27; -0,04)]. Este estudo demonstrou a relação da frequência cardíaca de repouso elevada com aumento das pressões arteriais sistólica e diastólica em ambos os sexos e com pressão de pulso em meninos, mesmo após controle para potenciais fatores de confusão, como obesidade geral e abdominal.

Epidemiological surveillance of tegumentary leishmaniasis: local territorial analysis.

PubMed

Soares, Valdenir Bandeira; Almeida, Andréa Sobral de; Sabroza, Paulo Chagastelles; Vargas, Waldemir Paixão

2017-06-26

To propose a new operational unit in the locality scale capable of subsidizing the construction of an information system to control the transmission of tegumentary leishmaniasis at this scale, in a region of high endemicity of the Atlantic Forest. We examined the adequacy of data and instruments in an area of high endemicity in the Atlantic Forest located in the South of the State of Rio de Janeiro from 1990 to 2012. The study proposed an operational unit called Local Surveillance Unit to make all used databases compatible by adjusting census sectors. This enabled the overlap and comparison of information in different periods. The spreading process of the transmission of tegumentary leishmaniasis in the Baía da Ilha Grande region does not depend on great population movements, and can occur in areas with population growth or decrease. The data information system allowed the adequate identification and characterization of the place of residence. We identified relevant characteristics of the place of transmission, such as self-limited in time and not associated with recent deforestation. The results also highlight the lack of synchronicity in the case production in territorial units involved in the endemic-epidemic process, noting that this process is in constant motion. The transmission process seems more connected to the presence and movement of rodents that move continuously in the region than to the local density of vectors or the permanence of infected dogs at home. New control strategies targeted at the foci of transmission must be considered. The construction of a new operational unit, called Local Surveillance Unit, was instrumental in the endemic-epidemic process analysis. Propor uma nova unidade operacional na escala de localidade capaz de subsidiar a construção de um sistema de informação orientado para o controle da transmissão da leishmaniose tegumentar nesse nível. Uma região de alta endemicidade da Mata Atlântica no sul do estado do Rio de Janeiro de 1990 a 2012 foi selecionada para analisar a adequação dos dados e instrumentos. Uma unidade operacional denominada Unidade de Vigilância Local foi proposta para compatibilizar todos os bancos de dados utilizados por meio de ajustes dos setores censitários. Isso possibilitou a sobreposição das informações e a sua comparação em diferentes períodos. O processo de deslocamento da transmissão da leishmaniose tegumentar na região da Baía da Ilha Grande não dependeu de movimentos populacionais importantes, podendo ocorrer tanto em áreas com crescimento como em áreas com decremento populacional. Os dados do sistema de informação permitiram a identificação e caracterização adequada do local de residência. Identificaram-se características relevantes do lugar de transmissão, como autolimitados no tempo e não associados a desmatamentos recentes. Os resultados evidenciam também a falta de sincronicidade na produção de casos nas unidades territoriais envolvidas no processo endêmico-epidêmico, mostrando que esse processo está em constante movimento. O processo de transmissão parece estar mais ligado à presença e circulação de roedores que se desloquem continuadamente na região do que da densidade local de vetores ou da permanência de cães infectados no domicílio. Novas estratégias de controle orientadas para os focos de transmissão devem ser consideradas. A construção de uma nova unidade operacional, denominada Unidade de Vigilância Local, foi fundamental na análise do processo endêmico-epidêmico.

Prediction of seasonal water-table fluctuations in La Pampa and Buenos Aires, Argentina

NASA Astrophysics Data System (ADS)

Tanco, Raúl; Kruse, Eduardo

2001-07-01

The fluctuation of the water table east of La Pampa province and northwest of Buenos Aires province, Argentina, influences agricultural production in the region because it is closely related to the alternation of dry and wet periods. Sea-surface temperature (SST) anomalies have been used as predictors to forecast atmospheric variables in different regions of the world. The objective of this work is to present a simple model to forecast seasonal rainfall using SST distribution in the Pacific Ocean as a predictor. Once the relationship between precipitation and water-table fluctuations was established, a methodology for the prediction of water-table fluctuations was developed. A good agreement between observed and predicted water-table fluctuations was found when estimating water-table fluctuations in the summer and autumn seasons. Résumé. Les fluctuations de la nappe à l'est de la province de La Pampa et au nord-ouest de la province de Buenos Aires (Argentine) influence la production agricole de la région parce qu'elle est étroitement liée à l'alternance de saisons sèches et humides. Les anomalies de la température de surface de l'océan (SST) ont été utilisées comme prédicteurs pour prévoir les variables atmosphériques dans différentes régions du monde. L'objectif de ce travail est de présenter un modèle simple de prévision des précipitations saisonnières en utilisant comme prédicteur la distribution des SST dans l'Océan Pacifique. Une fois que la relation entre les fluctuations des précipitations et celles de la nappe a été établie, une méthodologie de prédiction des variations de la nappe a été mise au point. Un bon accord entre les variations de la nappe observées et celles prédites a été trouvé pour les estimations des variations de nappe en été et en automne. Resumen. La fluctuación del nivel freático al este de la provincia de La Pampa y al nordeste de la de Buenos Aires (Argentina) repercute en la producción agrícola de la región, ya que está íntimamente relacionada con la alternancia de períodos secos y húmedos. Se ha utilizado las anomalías de la temperatura superficial del mar (TSM) para predecir las variables atmosféricas en diferentes áreas del mundo. El objetivo de este trabajo es presentar un modelo sencillo para pronosticar la precipitación estacional por medio de la distribución de TSM en el Océano Pacífico. Una vez establecida la relación entre la precipitación y las fluctuaciones del nivel freático, se desarrolló una metodología para predecir las fluctuaciones de éste. Se obtuvo un buen ajuste entre las fluctuaciones predichas y observadas del nivel freático en las estaciones de verano y otoño.

Hypertriglyceridemic Waist Phenotype and Changes in the Fasting Glycemia and Blood Pressure in Children and Adolescents Over One-Year Follow-Up Period.

PubMed

Costa, Priscila Ribas de Farias; Assis, Ana Marlúcia Oliveira; Cunha, Carla de Magalhães; Pereira, Emile Miranda; Jesus, Gabriela Dos Santos de; Silva, Lais Eloy Machado da; Alves, Wilanne Pinheiro de Oliveira

2017-07-01

The hypertriglyceridemic waist (HTW) phenotype is defined as the simultaneous presence of increased waist circumference (WC) and serum triglycerides (TG) levels and it has been associated with cardiometabolic risk in children and adolescents. The objective was to evaluate the influence of HTW phenotype in the fasting glycemia and blood pressure in children and adolescents over one-year follow-up period. It is a cohort study involving 492 children and adolescents from 7 to 15 years old, both genders, who were submitted to anthropometric, biochemical and clinical evaluation at the baseline, and also after 6 and 12 months of follow-up. Generalized Estimating Equation (GEE) models were calculated to evaluate the longitudinal influence of the HTW phenotype in the glycemia and blood pressure over one-year. It was observed a prevalence of 10.6% (n = 52) of HTW phenotype in the students. The GEE models identified that students with HTW phenotype had an increase of 3.87 mg/dl in the fasting glycemia mean (CI: 1.68-6.05) and of 3.67mmHg in the systolic blood pressure (SBP) mean (CI: 1.55-6.08) over one-year follow-up, after adjusting for confounding variables. The results of this study suggest that HTW phenotype is a risk factor for longitudinal changes in glycemia and SBP in children and adolescents over one-year follow-up period. O fenótipo de cintura hipertrigliceridêmica (CHT) é definido como a presença simultânea de circunferência de cintura (CC) e níveis séricos de triglicérides (TG) aumentados e tem sido associado com risco cardiometabólico em crianças e adolescentes. Avaliar a influência do fenótipo CHT na glicemia de jejum e na pressão arterial em crianças e adolescentes em um período de acompanhamento de um ano. Trata-se de um estudo de coorte envolvendo 492 crianças e adolescentes de 7 a 15 anos de ambos os sexos, que foram submetidos à avaliação antropométrica, bioquímica e clínica no início e também após 6 e 12 meses de seguimento. Os modelos de Equação de Estimulação Generalizada (GEE) foram calculados para avaliar a influência longitudinal do fenótipo CHT na glicemia e na pressão arterial ao longo de um ano. Foi observada uma prevalência de 10,6% (n = 52) do fenótipo CHT nos estudantes. Os modelos GEE identificaram que os estudantes com fenótipo CHT apresentaram aumento de 3,87 mg/dl na média de glicemia em jejum (IC: 1,68-6,05) e de 3,67 mmHg na pressão arterial sistólica media (PAS) (IC: 1,55-6,08) depois de um ano de acompanhamento, após ajuste para variáveis de confusão. Os resultados deste estudo sugerem que o fenótipo CHT é um fator de risco para alterações longitudinais da glicemia e da PAS em crianças e adolescentes em um período de um ano de seguimento.

Impact of Chloroquine on Viral Load in Breast Milk

PubMed Central

Semrau, Katherine; Kuhn, Louise; Kasonde, Prisca; Sinkala, Moses; Kankasa, Chipepo; Shutes, Erin; Vwalika, Cheswa; Ghosh, Mrinal; Aldrovandi, Grace; Thea, Donald M.

2006-01-01

Summary The anti-malarial agent chloroquine has activity against HIV. We compared the effect of chloroquine (n = 18) to an anti-malarial agent without known anti-HIV-activity, sulfadoxine-pyrimethamine (n = 12), on breast milk HIV RNA levels among HIV-infected breastfeeding women in Zambia. After adjusting for CD4 count and plasma viral load, chloroquine was associated with a trend towards lower levels of HIV RNA in breast milk compared with sulfadoxine-pyrimethamine (P 0.05). Higher breastmilk viral load was also observed among women receiving presumptive treatment = for symptomatic malaria compared with asymptomatic controls and among controls reporting fever in the prior week. Further research is needed to determine the potential role of chloroquine in prevention of HIV transmission through breastfeeding. Impacte de la chloroquine sur la charge virale dans le lait maternelle La chloroquine, agent antimalarique, a une activité contre le VIH. Nous avons comparé l’effet de la chloroquine à celui d’un autre agent antimalarique, la sulfadoxine-pyrimethamine, dont l’activité sur le VIH n’est pas connue, en mesurant les taux d’ARN de VIH dans le lait maternel de femmes allaitantes infectées par le VIH en Zambie. Après ajustement pour les taux de CD4 et la charge virale dans le plasma, la chloroquine comparée à la sulfadoxine pyrimethamine était associée à une tendance vers des teneurs plus bas en ARN de VIH dans le lait maternel (P = 0,05). Des charges virales plus élevées dans le lait maternel étaient aussi observées chez des femmes recevant un traitement présomptif pour des symptômes de malaria par rapport aux contrôles asymptomatiques et par rapport à des contrôles rapportant de la fièvre durant la première semaine. Des études supplémentaires sont nécessaires pour déterminer le rôle potentiel de la chloroquine dans la prévention de la transmission du VIH par l’allaitement maternel. mots clésVIH, malaria, allaitement maternel, chloroquine, sulfadoxine-pyrimethamine, charge virale du lait maternel, fièvre Impacto de la cloroquina en la carga viral de la leche materna El antimalárico cloroquina tiene actividad frente al VIH. Comparamos el efecto de la cloroquina (n = 18) frente a un antimalárico sin actividad anti-VIH conocida, la sulfadoxina-pirimetamina (n = 12), en los niveles de ARN en la leche materna de mujeres infectadas con VIH, en Zambia. Después de ajustar para recuento de CD4 y la carga viral en plasma, se asoció a la cloroquina con una tendencia hacia menores niveles de ARN del VIH en leche materna, comparado con la sulfadoxina pirimetamina (P = 0.05). También se observó una mayor carga viral en la leche materna de mujeres recibiendo tratamiento presuntivo para malaria sintomática, que en los controles asintomáticos y controles que habáan reportado fiebre la semana anterior. Es necesario realizar más estudios para determinar el papel potencial de la cloroquina en la prevención de la trasmisión de VIH a través de la lactancia materna. palabras claveVIH, malaria, lactancia materna, cloroquina, sulfadoxina pirimetamina, transmisión vertical, leche materna, carga viral, fiebre PMID:16772000

Behaviour of the pH Adjustment, Ion Exchange and Concentrate Precipitation Stages in the Acid Leaching of Uranium Phosphate Ores; TRATAMIENTO DE DISOLUCIONES DE LIXIVIACION DE MINERALES DE URANIO EN PRESENCIA DE FOSFATOS. COMPORTAMIENTO EN LAS ETAPAS DE AJUSTE DE PH, CAMBIO DE ION Y PRECIPITACION DE CONCENTRADOS (in Spanish)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aguilar, J.E.; Hueda, A.U.

Recovery of U from acid leach solutions of phosphate ore was studied. It was found that predictions can be made concerning solids removal and U recovery in the pH adjustment stage, resin U capacity, eluating agent suitability, ion exchange stage eluation velocity and eluate U concentration, and composition of the precipitate formed in the concentration stage. The results are valid in the concentration range 0.3 to 0. 8 g U/sub 3/O/sub 8//1. (J.R.D.)

Cost Structure and Life Cycle Cost (LCC) for Military Systems (structures de couts globaux de possession (LCC) pour systemes militaires)

DTIC Science & Technology

2003-06-01

variables. Dans le plan (p = 2), la méthode de régression linéaire ajuste au nuage de points une droite qui minimise la somme des écarts au carré entre...cohérence de la définition du produit. Par exemple, pour un blindé, il faut valider l ’ « harmonie » du trio masse, puissance du moteur et vitesse maximale. Il...en condition opérationnelle. De façon plus générale, la DGA devrait s’orienter vers l ‘élaboration et l’acquisition de modèle d’estimation de coûts du

Study of continuous precipitation of ADU for the implantation in the pilot installation at the Atomic Energy Institute, Brazil. Estudo de precipitacao continua de DUA para a implantacao na instalacao piloto CEQ-IEA (in Portuguese)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Araujo, Jose Adroalado de

1974-05-15

The paper deals with the ammonium diuranate continuous precipitation with a high chemical purity degree from uranyl nitrate solutions, using 1.2 and 2.4 ammonium hydroxide solutions and gaseous NH{sub 3} as a precipitating agent. The precipitations were carried out in a continuous procedure with one and two stages. The variables studied were the NH[sub 4}OH solutions concentration, ADU precipitation curve, the flow rate of reactants, the temperature of the precipitation, pH of the suspended ADU, and ammonium diuranate filtrability. The experimental work performed and the data obtained permitted the design of a chemical reactor for the continuous nuclear grade ADUmore » precipitation at the Chemical Engineering Department of the Atomic Energy Institute of Sao Paulo.« less

Weathering resistance of thin plasma polymer films on pre-coated steel =

NASA Astrophysics Data System (ADS)

Serra, Ricardo Gil Henriques

O trabalho apresentado teve origem no projecto de investigacao “Tailored Thin Plasma Polymers Films for Surface Engineering of Coil Coated Steel”, financiado pelo Programa Europeu ECSC Steel Research. Sistemas de aco galvanizado pre-pintado em banda a base de poliester e poliuretano foram submetidos a um processo de polimerizacao por plasma onde um filme fino foi depositado de modo a modificar as propriedades de superficie. Foram usados reactores de catodo oco, microondas e radio frequencia para a deposicao do polimero fino. Os sistemas preparados foram analisados de modo a verificar a influencia do processo de polimerizacao por plasma na alteracao das propriedades barreira dos sistemas pre-pintados em banda. Foi estudado o efeito dos diferentes passos do processo de polimerizacao por plasma, bem como o efeito de diferentes variaveis operatorias. A mistura precursora foi variada de modo a modificar as propriedades da superficie de modo a poder vir a obter maior hidrofobicidade, maior resistencia a marcas digitais, bem como maior facilidade de limpeza. Os testes foram conduzidos em solucao de NaCl 0,5 M. Para o trabalho foram usadas tecnicas de analise da morfologia da superficie como Microscopia de Forca Atomica e Microscopia Electronica de Varrimento. As propriedades electroquimicas dos sistemas foram estudadas por Espectroscopia de Impedancia Electroquimica. A estrutura dos filmes gerados no processo de polimerizacao por plasma foi caracterizada por Microscopia de Transmissao Electronica. A modificacao das propriedades opticas devido ao processo de polimerizacao por plasma foi tambem obtida.

Working hours and health in nurses of public hospitals according to gender.

PubMed

Fernandes, Juliana da Costa; Portela, Luciana Fernandes; Griep, Rosane Härter; Rotenberg, Lúcia

2017-06-26

To assess the association between weekly working hours and self-rated health of nurses in public hospitals in Rio de Janeiro, State of Rio de Janeiro, Brazil. A total of 3,229 nurses (82.7% of the eligible group) participated in this cross-sectional study, carried out between April 2010 and December 2011. The collection instrument consisted of a self-administered multidimensional questionnaire. The weekly working hours were calculated from a recall of the daily hours worked over seven consecutive days; this variable was categorized according to tertiles of distribution for men and women. The outcome of interest, self-rated health, was categorized into three levels: good (very good and good), regular, and poor (poor and very poor). The statistical analysis of the data included bivariate and multivariate analyses, having as reference group those with short working hours (first tertile). All the analyses were stratified by gender and elaborated using the program SPSS. Among women, the group corresponding to the longest working week (more than 60.5 hours per week) were more likely to report regular self-rated health, compared with those with shorter working hours, after adjusting for confounding factors (OR = 1.30; 95%CI 1.02-1.67). Among men, those with average working hours (49.5-70.5 hours per week) were more than twice as likely to rate their health as regular (OR = 2.17; 95%CI 1.08-4.35) compared to those with shorter working hours (up to 49.5 hours). There was no significant association between long working hours and poor self-rated health. The results point to the urgent need to promote interventions in the organization of work and appreciation of the nursing profession, in order to reduce the number of multiple jobs and thus contribute to mitigate potential effects on the health of workers and the quality of care in hospitals. Avaliar a associação entre horas de trabalho semanais e autoavaliação de saúde de enfermeiros em hospitais públicos do Rio de Janeiro, RJ, Brasil. Um total de 3.229 enfermeiros (82,7% do grupo de elegíveis) participou deste estudo transversal, realizado entre abril de 2010 e dezembro de 2011. O instrumento de coleta consistiu em um questionário multidimensional autopreenchido. As horas de trabalho semanais foram calculadas a partir de um recordatório das horas diárias de trabalho ao longo de sete dias consecutivos; esta variável foi categorizada de acordo com tercis da distribuição para homens e mulheres. O desfecho de interesse, auto-avaliação de saúde, foi categorizado em três níveis: bom (muito bom e bom), regular e ruim (ruim e muito ruim). A análise estatística dos dados incluiu análises bivaridas e multivariadas, tendo como grupo de referência aqueles com jornadas curtas de trabalho (primeiro tercil). Todas as análises foram estratificadas por sexo e elaboradas no programa SPSS. Entre as mulheres, o grupo correspondente à semana de trabalho mais longa (mais de 60,5 horas por semana) tinha maior probabilidade de relatar autoavaliação de saúde como regular, em comparação com aqueles com jornada curta, após o ajuste para fatores de confusão (OR = 1,30; IC95% 1,02-1,67). Entre os homens, aqueles com jornada média (49,5-70,5 horas por semana) tiveram mais que o dobro da probabilidade de avaliar sua saúde como regular (OR = 2,17; IC95% 1,08-4,35) em comparação com aqueles com a semana de trabalho mais curta (até 49,5 horas). Não houve associação significativa entre longas horas de trabalho e autoavaliação de saúde ruim. Os resultados apresentados apontam para a urgência em promover intervenções na organização do trabalho e valorização da profissão de enfermagem, de modo a reduzir o múltiplo vínculo e assim contribuir para mitigar possíveis efeitos sobre a saúde dos trabalhadores e a qualidade do atendimento nos hospitais.

Propriétés optiques du silicium mésoporeux et ses applications potentielles

NASA Astrophysics Data System (ADS)

Lévy-Clément, C.; Bastide, S.

2002-04-01

Une couche de silicium poreux formée électrochimiquement à la surface de l'émetteur d'une jonction p-n+ peut jouer efficacement le rôle d'une couche antireflet pour les cellules solaires au silicium. En ajustant la densité de courant et le temps de formation, des simple et double-couches antireflet aux propriétés optiques optimisées ont pu être réalisées, conduisant à des réflectivités effectives respectives de 7.3 et 2.7 % sur l'ensemble du spectre solaire utile pour les cellules (400-1000 nm). De hauts rendements de conversion de l'énergie solaire, de l'ordre de 13-13.4 %, sont ainsi actuellement obtenus pour des cellules au silicium multicristallin avec une couche antireflet de silicium poreux.

Socioeconomic determinants of access to health services among older adults: a systematic review.

PubMed

Almeida, Ana Paula Santana Coelho; Nunes, Bruno Pereira; Duro, Suele Manjourany Silva; Facchini, Luiz Augusto

2017-05-15

The objective of this study was to analyze the association between the socioeconomic characteristics and the access to or use of health services among older adults. This is a systematic review of the literature. The search has been carried out in the databases PubMed, LILACS and Web of Science, without restriction of dates and languages; however we have included only articles published in Portuguese, English, and Spanish. The inclusion criteria were: observational design, socioeconomic factors as variables of interest in the analysis of the access to or use of health services among older adults, representative sample of the target population, adjustment for confounding factors, and no selection bias. We have found 5,096 articles after deleting duplicates and 36 of them have been selected for review after the process of reading and evaluating the inclusion criteria. Higher income and education have been associated with the use and access to medical appointments in developing countries and some developed countries. The same association has been observed in dental appointments in all countries. Most studies have shown no association between socioeconomic characteristics and the use of inpatient and emergency services. We have identified greater use of home visits in lower-income individuals, with the exception of the United States. We have observed an unequal access to or use of health services in most countries, varying according to the type of service used. The expansion of the health care coverage is necessary to reduce this unequal access generated by social inequities. Analisar a associação entre características socioeconômicas e acesso ou utilização de serviços de saúde entre idosos. Revisão sistemática da literatura. A busca foi realizada nas bases de dados PubMed, Lilacs e Web of Science, sem restrição de datas e idiomas, entretanto foram incluídos somente os artigos publicados em português, inglês e espanhol. Foram critérios de inclusão: ter delineamento observacional; possuir os fatores socioeconômicos como variáveis de interesse na análise do acesso ou utilização de serviços de saúde entre idosos; ter amostra representativa da população alvo; fazer ajuste para fatores de confusão; e não apresentar viés de seleção. Foram encontrados 5.096 artigos após a exclusão de duplicidades e 36 foram selecionados para a revisão após o processo de leitura e avaliação dos critérios de inclusão. Maior renda e escolaridade estiveram associadas à utilização e acesso a consultas médicas nos países em desenvolvimento e em alguns países desenvolvidos. A mesma associação foi observada nas consultas odontológicas em todos os países. A maioria dos estudos não apresentou associação entre características socioeconômicas e uso de serviços de internação e emergência. Foi identificado maior uso de visita domiciliar em indivíduos de menor renda, com exceção dos Estados Unidos. Observou-se desigualdade no acesso ou na utilização de serviços de saúde na maior parte dos países, variando em relação ao tipo de serviço utilizado. A ampliação da cobertura de serviços de saúde faz-se necessária para a redução da desigualdade no acesso gerada por iniquidades sociais.

JPRS Report, Science & Technology, China: Energy.

DTIC Science & Technology

1992-03-30

breeder reactors should become...the primary type of reactors . In developing breeder reactors , we should follow the path of using metal fuel. Breeder reactors give us more time to...first reactor used for power generation was a fast reactor : the " Breeder 1" reactor at the Idaho National Reactor Test Center which was used to

Biofilm reactors for industrial bioconversion processes: employing potential of enhanced reaction rates

PubMed Central

Qureshi, Nasib; Annous, Bassam A; Ezeji, Thaddeus C; Karcher, Patrick; Maddox, Ian S

2005-01-01

This article describes the use of biofilm reactors for the production of various chemicals by fermentation and wastewater treatment. Biofilm formation is a natural process where microbial cells attach to the support (adsorbent) or form flocs/aggregates (also called granules) without use of chemicals and form thick layers of cells known as "biofilms." As a result of biofilm formation, cell densities in the reactor increase and cell concentrations as high as 74 gL-1 can be achieved. The reactor configurations can be as simple as a batch reactor, continuous stirred tank reactor (CSTR), packed bed reactor (PBR), fluidized bed reactor (FBR), airlift reactor (ALR), upflow anaerobic sludge blanket (UASB) reactor, or any other suitable configuration. In UASB granular biofilm particles are used. This article demonstrates that reactor productivities in these reactors have been superior to any other reactor types. This article describes production of ethanol, butanol, lactic acid, acetic acid/vinegar, succinic acid, and fumaric acid in addition to wastewater treatment in the biofilm reactors. As the title suggests, biofilm reactors have high potential to be employed in biotechnology/bioconversion industry for viable economic reasons. In this article, various reactor types have been compared for the above bioconversion processes. PMID:16122390

Le Programme d’enseignement décentralisé et la rétention des médecins omnipraticiens dans la région du Bas-Saint-Laurent au Québec

PubMed Central

Bustinza, Ray; Gagnon, Suzanne; Burigusa, Guillaume

2009-01-01

Résumé OBJECTIF Le but de cette étude était d’évaluer l’impact sur la rétention des médecins omnipraticiens du programme d’enseignement décentralisé, des mesures incitatives financières et de l’origine du médecin. DEVIS Nos données sont tirées de la banque du suivi des effectifs médicaux de la Régie régionale de la santé et des services sociaux du Bas-Saint-Laurent. Les questionnaires complétés par les médecins dans le cadre de cette étude ont été envoyés par la poste. MILIEU Bas-Saint-Laurent, Qué. PARTICIPANTS Des médecins omnipraticiens pratiquant depuis 1985 à août 2003 dans la région. MÉTHODOLOGIE Pour les analyses statistiques, nous avons fait appel à l’analyse multivariée, en utilisant le modèle des taux proportionnels de Cox de l’analyse de survie. RÉSULTATS La probabilité ajustée de demeurer en région, selon l’exposition aux stages de résidence en région, présente une tendance positive avec un rapport de 2,12 (P = 0,15). La probabilité de rester dans la région grimpe jusqu’à 4,5 fois plus (P < 0,01) pour un médecin originaire du Bas-Saint-Laurent. Le lien avec les mesures incitatives financières montre que la probabilité ajustée de demeurer dans le Bas-Saint-Laurent n’est pas significativement différente selon qu’ils aient ou non reçu les primes d’installation ou la bourse de la Régie de l’assurance-maladie du Québec. CONCLUSION Les facteurs les plus prometteurs pour retenir les médecins omnipraticiens en région touchent le recrutement en médecine de candidats des régions rurales et la formation décentralisée. PMID:19752242

Control of reactor coolant flow path during reactor decay heat removal

DOEpatents

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source spectrum of the NBSR reactor at the NIST Center for Neutron Research

NASA Astrophysics Data System (ADS)

Cook, J. C.; Barker, J. G.; Rowe, J. M.; Williams, R. E.; Gagnon, C.; Lindstrom, R. M.; Ibberson, R. M.; Neumann, D. A.

2015-08-01

The recent expansion of the National Institute of Standards and Technology (NIST) Center for Neutron Research facility has offered a rare opportunity to perform an accurate measurement of the cold neutron spectrum at the exit of a newly-installed neutron guide. Using a combination of a neutron time-of-flight measurement, a gold foil activation measurement, and Monte Carlo simulation of the neutron guide transmission, we obtain the most reliable experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source brightness to date. Time-of-flight measurements were performed at three distinct fuel burnup intervals, including one immediately following reactor startup. Prior to the latter measurement, the hydrogen was maintained in a liquefied state for an extended period in an attempt to observe an initial radiation-induced increase of the ortho (o)-hydrogen fraction. Since para (p)-hydrogen has a small scattering cross-section for neutron energies below 15 meV (neutron wavelengths greater than about 2.3 Å), changes in the o- p hydrogen ratio and in the void distribution in the boiling hydrogen influence the spectral distribution. The nature of such changes is simulated with a continuous-energy, Monte Carlo radiation-transport code using 20 K o and p hydrogen scattering kernels and an estimated hydrogen density distribution derived from an analysis of localized heat loads. A comparison of the transport calculations with the mean brightness function resulting from the three measurements suggests an overall o- p ratio of about 17.5(±1) % o- 82.5% p for neutron energies<15 meV, a significantly lower ortho concentration than previously assumed.

Asymmetric reduction of ketones and β-keto esters by (S)-1-phenylethanol dehydrogenase from denitrifying bacterium Aromatoleum aromaticum.

PubMed

Dudzik, A; Snoch, W; Borowiecki, P; Opalinska-Piskorz, J; Witko, M; Heider, J; Szaleniec, M

2015-06-01

Enzyme-catalyzed enantioselective reductions of ketones and keto esters have become popular for the production of homochiral building blocks which are valuable synthons for the preparation of biologically active compounds at industrial scale. Among many kinds of biocatalysts, dehydrogenases/reductases from various microorganisms have been used to prepare optically pure enantiomers from carbonyl compounds. (S)-1-phenylethanol dehydrogenase (PEDH) was found in the denitrifying bacterium Aromatoleum aromaticum (strain EbN1) and belongs to the short-chain dehydrogenase/reductase family. It catalyzes the stereospecific oxidation of (S)-1-phenylethanol to acetophenone during anaerobic ethylbenzene mineralization, but also the reverse reaction, i.e., NADH-dependent enantioselective reduction of acetophenone to (S)-1-phenylethanol. In this work, we present the application of PEDH for asymmetric reduction of 42 prochiral ketones and 11 β-keto esters to enantiopure secondary alcohols. The high enantioselectivity of the reaction is explained by docking experiments and analysis of the interaction and binding energies of the theoretical enzyme-substrate complexes leading to the respective (S)- or (R)-alcohols. The conversions were carried out in a batch reactor using Escherichia coli cells with heterologously produced PEDH as whole-cell catalysts and isopropanol as reaction solvent and cosubstrate for NADH recovery. Ketones were converted to the respective secondary alcohols with excellent enantiomeric excesses and high productivities. Moreover, the progress of product formation was studied for nine para-substituted acetophenone derivatives and described by neural network models, which allow to predict reactor behavior and provides insight on enzyme reactivity. Finally, equilibrium constants for conversion of these substrates were derived from the progress curves of the reactions. The obtained values matched very well with theoretical predictions.

Neutron fluxes in test reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youinou, Gilles Jean-Michel

Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

Reactor engineering support of operations at the Davis-Besse nuclear power station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelley, D.B.

1995-12-31

Reactor engineering functions differ greatly from unit to unit; however, direct support of the reactor operators during reactor startups and operational transients is common to all units. This paper summarizes the support the reactor engineers provide the reactor operators during reactor startups and power changes through the use of automated computer programs at the Davis-Besse nuclear power station.

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2011 CFR

2011-01-01

... chapter by the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director...

Nuclear Reactors. Revised.

ERIC Educational Resources Information Center

Hogerton, John F.

This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2012 CFR

2012-01-01

... chapter by the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director...

Nuclear reactor construction with bottom supported reactor vessel

DOEpatents

Sharbaugh, John E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. Kokkinos

2005-04-28

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophymore » on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.« less

Heat transfer analysis of cylindrical anaerobic reactors with different sizes: a heat transfer model.

PubMed

Liu, Jiawei; Zhou, Xingqiu; Wu, Jiangdong; Gao, Wen; Qian, Xu

2017-10-01

The temperature is the essential factor that influences the efficiency of anaerobic reactors. During the operation of the anaerobic reactor, the fluctuations of ambient temperature can cause a change in the internal temperature of the reactor. Therefore, insulation and heating measures are often used to maintain anaerobic reactor's internal temperature. In this paper, a simplified heat transfer model was developed to study heat transfer between cylindrical anaerobic reactors and their surroundings. Three cylindrical reactors of different sizes were studied, and the internal relations between ambient temperature, thickness of insulation, and temperature fluctuations of the reactors were obtained at different reactor sizes. The model was calibrated by a sensitivity analysis, and the calibrated model was well able to predict reactor temperature. The Nash-Sutcliffe model efficiency coefficient was used to assess the predictive power of heat transfer models. The Nash coefficients of the three reactors were 0.76, 0.60, and 0.45, respectively. The model can provide reference for the thermal insulation design of cylindrical anaerobic reactors.

Solvent refined coal reactor quench system

DOEpatents

Thorogood, Robert M.

1983-01-01

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

Solvent refined coal reactor quench system

DOEpatents

Thorogood, R.M.

1983-11-08

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

Nuclear reactor neutron shielding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Speaker, Daniel P; Neeley, Gary W; Inman, James B

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactormore » cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.« less

Reactor pressure vessel head vents and methods of using the same

DOEpatents

Gels, John L; Keck, David J; Deaver, Gerald A

2014-10-28

Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belles, Randy; Poore, III, Willis P.; Brown, Nicholas R.

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-basedmore » description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.« less

10 CFR 52.167 - Issuance of manufacturing license.

Code of Federal Regulations, 2010 CFR

2010-01-01

... proposed reactor(s) can be incorporated into a nuclear power plant and operated at sites having... design and manufacture the proposed nuclear power reactor(s); (5) The proposed inspections, tests... the construction of a nuclear power facility using the manufactured reactor(s). (2) A holder of a...

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

Bhattacharyya, Samit K.; Lipinski, Walter C.; Hanan, Nelson A.

A novel plant protection system designed for use in the TREAT Upgrade (TU) reactor is described. The TU reactor is designed for controlled transient operation in the testing of reactor fuel behavior under simulated reactor accident conditions. Safe operation of the reactor is of paramount importance and the Plant Protection System (PPS) had to be designed to exacting requirements. Researchers believe that the strategy developed for the TU has potential application to the multimegawatt space reactors and represents the state of the art in terrestrial transient reactor protection systems.

Process and apparatus for adding and removing particles from pressurized reactors

DOEpatents

Milligan, John D.

1983-01-01

A method for adding and removing fine particles from a pressurized reactor is provided, which comprises connecting the reactor to a container, sealing the container from the reactor, filling the container with particles and a liquid material compatible with the reactants, pressurizing the container to substantially the reactor pressure, removing the seal between the reactor and the container, permitting particles to fall into or out of the reactor, and resealing the container from the reactor. An apparatus for adding and removing particles is also disclosed.

Effects of imperfect mixing on low-density polyethylene reactor dynamics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Villa, C.M.; Dihora, J.O.; Ray, W.H.

1998-07-01

Earlier work considered the effect of feed conditions and controller configuration on the runaway behavior of LDPE autoclave reactors assuming a perfectly mixed reactor. This study provides additional insight on the dynamics of such reactors by using an imperfectly mixed reactor model and bifurcation analysis to show the changes in the stability region when there is imperfect macroscale mixing. The presence of imperfect mixing substantially increases the range of stable operation of the reactor and makes the process much easier to control than for a perfectly mixed reactor. The results of model analysis and simulations are used to identify somemore » of the conditions that lead to unstable reactor behavior and to suggest ways to avoid reactor runaway or reactor extinction during grade transitions and other process operation disturbances.« less

Startup of reactors for anoxic ammonium oxidation: experiences from the first full-scale anammox reactor in Rotterdam.

PubMed

van der Star, Wouter R L; Abma, Wiebe R; Blommers, Dennis; Mulder, Jan-Willem; Tokutomi, Takaaki; Strous, Marc; Picioreanu, Cristian; van Loosdrecht, Mark C M

2007-10-01

The first full-scale anammox reactor in the world was started in Rotterdam (NL). The reactor was scaled-up directly from laboratory-scale to full-scale and treats up to 750 kg-N/d. In the initial phase of the startup, anammox conversions could not be identified by traditional methods, but quantitative PCR proved to be a reliable indicator for growth of the anammox population, indicating an anammox doubling time of 10-12 days. The experience gained during this first startup in combination with the availability of seed sludge from this reactor, will lead to a faster startup of anammox reactors in the future. The anammox reactor type employed in Rotterdam was compared to other reactor types for the anammox process. Reactors with a high specific surface area like the granular sludge reactor employed in Rotterdam provide the highest volumetric loading rates. Mass transfer of nitrite into the biofilm is limiting the conversion of those reactor types that have a lower specific surface area. Now the first full-scale commercial anammox reactor is in operation, a consistent and descriptive nomenclature is suggested for reactors in which the anammox process is employed.

A small, 1400 K, reactor for Brayton space power systems.

NASA Technical Reports Server (NTRS)

Lantz, E.; Mayo, W.

1972-01-01

An investigation was conducted to determine minimum dimensions and minimum weight obtainable in a design for a reactor using uranium-233 nitride or plutonium-239 nitride as fuel. Such a reactor had been considered by Krasner et al. (1971). Present space power status is discussed, together with questions of reactor design and power distribution in the reactor. The characteristics of various reactor types are compared, giving attention also to a zirconium hydride reactor.

Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

NASA Technical Reports Server (NTRS)

Roth, R. J.

1976-01-01

The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

KINETICS OF TREAT USED AS A TEST REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickerman, C.E.; Johnson, R.D.; Gasidlo, J.

1962-05-01

An analysis is presented concerning the reactor kinetics of TREAT used as a pulsed, engineering test reactor for fast reactor fuel element studies. A description of the reactor performance is given for a wide range of conditions associated with its use as a test reactor. Supplemental information on meltdown experimentation is included. (J.R.D.)

Generating unstructured nuclear reactor core meshes in parallel

DOE PAGES

Jain, Rajeev; Tautges, Timothy J.

2014-10-24

Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor coremore » examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.« less

Nuclear reactor having a polyhedral primary shield and removable vessel insulation

DOEpatents

Ekeroth, Douglas E.; Orr, Richard

1993-01-01

A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel.

Nuclear reactor having a polyhedral primary shield and removable vessel insulation

DOEpatents

Ekeroth, D.E.; Orr, R.

1993-12-07

A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel. 5 figures.

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, C.W.

1985-02-19

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, Charles W.

1987-01-01

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Summary of NR Program Prometheus Efforts

DOE Office of Scientific and Technical Information (OSTI.GOV)

J Ashcroft; C Eshelman

2006-02-08

The Naval Reactors Program led work on the development of a reactor plant system for the Prometheus space reactor program. The work centered on a 200 kWe electric reactor plant with a 15-20 year mission applicable to nuclear electric propulsion (NEP). After a review of all reactor and energy conversion alternatives, a direct gas Brayton reactor plant was selected for further development. The work performed subsequent to this selection included preliminary nuclear reactor and reactor plant design, development of instrumentation and control techniques, modeling reactor plant operational features, development and testing of core and plant material options, and development ofmore » an overall project plan. Prior to restructuring of the program, substantial progress had been made on defining reference plant operating conditions, defining reactor mechanical, thermal and nuclear performance, understanding the capabilities and uncertainties provided by material alternatives, and planning non-nuclear and nuclear system testing. The mission requirements for the envisioned NEP missions cannot be accommodated with existing reactor technologies. Therefore concurrent design, development and testing would be needed to deliver a functional reactor system. Fuel and material performance beyond the current state of the art is needed. There is very little national infrastructure available for fast reactor nuclear testing and associated materials development and testing. Surface mission requirements may be different enough to warrant different reactor design approaches and development of a generic multi-purpose reactor requires substantial sacrifice in performance capability for each mission.« less

METHOD AND APPARATUS FOR CONTROLLING DIRECT-CYCLE NEUTRONIC REACTORS

DOEpatents

Reed, G.A.

1961-01-10

A control arrangement is offered for a boiling-water reactor. Boric acid is maintained in the water in the reactor and the amount in the reactor is controlled by continuously removing a portion of the water from the reactor, concentrating the boric acid by evaporating the water therefrom, returning a controlled amount of the acid to the reactor, and simultaneously controlling the water level by varying the rate of spent steam return to the reactor.

Control Means for Reactor

DOEpatents

Manley, J. H.

1961-06-27

An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

10 CFR 2.621 - Acceptance and docketing of application for early review of site suitability issues in a combined...

Code of Federal Regulations, 2014 CFR

2014-01-01

... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will inform the... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will accept for... New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they...

10 CFR 2.621 - Acceptance and docketing of application for early review of site suitability issues in a combined...

Code of Federal Regulations, 2013 CFR

2013-01-01

... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will inform the... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will accept for... New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they...

Low temperature pre-treatment of domestic sewage in an anaerobic hybrid or an anaerobic filter reactor.

PubMed

Elmitwalli, Tarek A; Sklyar, Vladimir; Zeeman, Grietje; Lettinga, Gatze

2002-05-01

The pre-treatment of domestic sewage for removal of suspended solids (SS) at a process temperature of 13 degrees C and an hydraulic retention time (HRT) of 4 h was investigated in an anaerobic filter (AF) and anaerobic hybrid (AH) reactor. The AF and the top of the AH reactor consisted of vertical sheets of reticulated polyurethane foam (RPF) with knobs. All biomass in the AF was only in attached form to avoid clogging and sludge washout. The AF reactor showed a significantly higher removal of total and suspended chemical oxygen demand (COD) than the AH reactor, respectively, 55% and 82% in the AF reactor and 34% and 53% in the AH reactor. Because the reactors were operated at a short HRT and low temperature, the hydrolysis, acidification and methanogenesis based on the influent COD were limited to, respectively, 12%, 21% and 23% for the AF reactor and 12%, 17% and 16% for the AH reactor. The excess sludge from the AH reactor was more stabilised and had a better settling capacity and dewaterability. However, the excess sludge from both the AH and AF reactors needed stabilisation. Therefore, the AF reactor is recommended for the pretreatment of domestic sewage at low temperatures.

Nuclear reactor cavity floor passive heat removal system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluidmore » communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.« less

Methods and apparatuses for deoxygenating pyrolysis oil

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baird, Lance Awender; Brandvold, Timothy A.; Frey, Stanley Joseph

Methods and apparatuses are provided for deoxygenating pyrolysis oil. A method includes contacting a pyrolysis oil with a deoxygenation catalyst in a first reactor at deoxygenation conditions to produce a first reactor effluent. The first reactor effluent has a first oxygen concentration and a first hydrogen concentration, based on hydrocarbons in the first reactor effluent, and the first reactor effluent includes an aromatic compound. The first reactor effluent is contacted with a dehydrogenation catalyst in a second reactor at conditions that deoxygenate the first reactor effluent while preserving the aromatic compound to produce a second reactor effluent. The second reactormore » effluent has a second oxygen concentration lower than the first oxygen concentration and a second hydrogen concentration that is equal to or lower than the first hydrogen concentration, where the second oxygen concentration and the second hydrogen concentration are based on the hydrocarbons in the second reactor effluent.« less

Methanation assembly using multiple reactors

DOEpatents

Jahnke, Fred C.; Parab, Sanjay C.

2007-07-24

A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

When Do Commercial Reactors Permanently Shut Down?

EIA Publications

2011-01-01

For those wishing to obtain current data, the following resources are available: U.S. reactors, go to the Energy Information Administration's nuclear reactor shutdown list. (Note: As of April 30, 2010, the last U.S. reactor to permanently shut down was Big Rock Point in 1997.) Foreign Reactors, go to the Power Reactor Information System (PRIS) on the International Atomic Energy Agency's website.

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 2.603 - Acceptance and docketing of application for early review of site suitability issues in a...

Code of Federal Regulations, 2013 CFR

2013-01-01

... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they are...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 2.603 - Acceptance and docketing of application for early review of site suitability issues in a...

Code of Federal Regulations, 2014 CFR

2014-01-01

... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they are...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cormon, S.; Fallot, M., E-mail: fallot@subatech.in2p3.fr; Bui, V.-M.

This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (ν{sup ¯}{sub e}) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of {sup 235}U, {supmore » 239}Pu and {sup 241}Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.« less

Bioaugmentation of activated sludge towards 3-chloroaniline removal with a mixed bacterial population carrying a degradative plasmid.

PubMed

Bathe, Stephan; Schwarzenbeck, Norbert; Hausner, Martina

2009-06-01

A bioaugmentation approach combining several strategies was applied to achieve degradation of 3-chloroaniline (3CA) in semicontinuous activated sludge reactors. In a first step, a 3CA-degrading Comamonas testosteroni strain carrying the degradative plasmid pNB2 was added to a biofilm reactor, and complete 3CA degradation together with spread of the plasmid within the indigenous biofilm population was achieved. A second set of reactors was then bioaugmented with either a suspension of biofilm cells removed from the carrier material or with biofilm-containing carrier material. 3CA degradation was established rapidly in all bioaugmented reactors, followed by a slow adaptation of the non-bioaugmented control reactors. In response to variations in 3CA concentration, all reactors exhibited temporary performance breakdowns. Whereas duplicates of the control reactors deviated in their behaviour, the bioaugmented reactors appeared more reproducible in their performance and population dynamics. Finally, the carrier-bioaugmented reactors showed an improved performance in the presence of high 3CA influent concentrations over the suspension-bioaugmented reactors. In contrast, degradation in one control reactor failed completely, but was rapidly established in the remaining control reactor.

A comparative study of thermophilic and mesophilic anaerobic co-digestion of food waste and wheat straw: Process stability and microbial community structure shifts.

PubMed

Shi, Xuchuan; Guo, Xianglin; Zuo, Jiane; Wang, Yajiao; Zhang, Mengyu

2018-05-01

Renewable energy recovery from organic solid waste via anaerobic digestion is a promising way to provide sustainable energy supply and eliminate environmental pollution. However, poor efficiency and operational problems hinder its wide application of anaerobic digestion. The effects of two key parameters, i.e. temperature and substrate characteristics on process stability and microbial community structure were studied using two lab-scale anaerobic reactors under thermophilic and mesophilic conditions. Both the reactors were fed with food waste (FW) and wheat straw (WS). The organic loading rates (OLRs) were maintained at a constant level of 3 kg VS/(m 3 ·d). Five different FW:WS substrate ratios were utilized in different operational phases. The synergetic effects of co-digestion improved the stability and performance of the reactors. When FW was mono-digested, both reactors were unstable. The mesophilic reactor eventually failed due to volatile fatty acid accumulation. The thermophilic reactor had better performance compared to mesophilic one. The biogas production rate of the thermophilic reactor was 4.9-14.8% higher than that of mesophilic reactor throughout the experiment. The shifts in microbial community structures throughout the experiment in both thermophilic and mesophilic reactors were investigated. With increasing FW proportions, bacteria belonging to the phylum Thermotogae became predominant in the thermophilic reactor, while the phylum Bacteroidetes was predominant in the mesophilic reactor. The genus Methanosarcina was the predominant methanogen in the thermophilic reactor, while the genus Methanothrix remained predominant in the mesophilic reactor. The methanogenesis pathway shifted from acetoclastic to hydrogenotrophic when the mesophilic reactor experienced perturbations. Moreover, the population of lignocellulose-degrading microorganisms in the thermophilic reactor was higher than those in mesophilic reactor, which explained the better performance of the thermophilic reactor. Copyright © 2018. Published by Elsevier Ltd.

Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

NASA Astrophysics Data System (ADS)

Wright, Steven A.; Houts, Michael

2001-02-01

Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

The chapter is devoted to the practical application of the fission process, mainly in nuclear reactors. After a historical discussion covering the natural reactors at Oklo and the first attempts to build artificial reactors, the fundamental principles of chain reactions are discussed. In this context chain reactions with fast and thermal neutrons are covered as well as the process of neutron moderation. Criticality concepts (fission factor η, criticality factor k) are discussed as well as reactor kinetics and the role of delayed neutrons. Examples of specific nuclear reactor types are presented briefly: research reactors (TRIGA and ILL High Flux Reactor), and some reactor types used to drive nuclear power stations (pressurized water reactor [PWR], boiling water reactor [BWR], Reaktor Bolshoi Moshchnosti Kanalny [RBMK], fast breeder reactor [FBR]). The new concept of the accelerator-driven systems (ADS) is presented. The principle of fission weapons is outlined. Finally, the nuclear fuel cycle is briefly covered from mining, chemical isolation of the fuel and preparation of the fuel elements to reprocessing the spent fuel and conditioning for deposit in a final repository.

Optimally moderated nuclear fission reactor and fuel source therefor

DOEpatents

Ougouag, Abderrafi M [Idaho Falls, ID; Terry, William K [Shelley, ID; Gougar, Hans D [Idaho Falls, ID

2008-07-22

An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

Flow rate analysis of wastewater inside reactor tanks on tofu wastewater treatment plant

NASA Astrophysics Data System (ADS)

Mamat; Sintawardani, N.; Astuti, J. T.; Nilawati, D.; Wulan, D. R.; Muchlis; Sriwuryandari, L.; Sembiring, T.; Jern, N. W.

2017-03-01

The research aimed to analyse the flow rate of the wastewater inside reactor tanks which were placed a number of bamboo cutting. The resistance of wastewater flow inside reactor tanks might not be occurred and produce biogas fuel optimally. Wastewater from eleven tofu factories was treated by multi-stages anaerobic process to reduce its organic pollutant and produce biogas. Biogas plant has six reactor tanks of which its capacity for waste water and gas dome was 18 m3 and 4.5 m3, respectively. Wastewater was pumped from collecting ponds to reactors by either serial or parallel way. Maximum pump capacity, head, and electrical motor power was 5m3/h, 50m, and 0.75HP, consecutively. Maximum pressure of biogas inside the reactor tanks was 55 mbar higher than atmosphere pressure. A number of 1,400 pieces of cutting bamboo at 50-60 mm diameter and 100 mm length were used as bacteria growth media inside each reactor tank, covering around 14,287 m2 bamboo area, and cross section area of inner reactor was 4,9 m2. In each reactor, a 6 inches PVC pipe was installed vertically as channel. When channels inside reactor were opened, flow rate of wastewater was 6x10-1 L.sec-1. Contrary, when channels were closed on the upper part, wastewater flow inside the first reactor affected and increased gas dome. Initially, wastewater flowed into each reactor by a gravity mode with head difference between the second and third reactor was 15x10-2m. However, head loss at the second reactor was equal to the third reactor by 8,422 x 10-4m. As result, wastewater flow at the second and third reactors were stagnant. To overcome the problem pump in each reactor should be installed in serial mode. In order to reach the output from the first reactor and the others would be equal, and biogas space was not filled by wastewater, therefore biogas production will be optimum.

Operators in the Plum Brook Reactor Facility Control Room

NASA Image and Video Library

1970-03-21

Donald Rhodes, left, and Clyde Greer, right, monitor the operation of the National Aeronautics and Space Administration’s (NASA) Plum Brook Reactor Facility from the control room. The 60-megawatt test reactor, NASA’s only reactor, was the eighth largest test reactor in the world. The facility was built by the Lewis Research Center in the late 1950s to study the effects of radiation on different materials that could be used to construct nuclear propulsion systems for aircraft or rockets. The reactor went critical for the first time in 1961. For the next two years, two operators were on duty 24 hours per day working on the fission process until the reactor reached its full-power level in 1963. Reactor Operators were responsible for monitoring and controlling the reactor systems. Once the reactor was running under normal operating conditions, the work was relatively uneventful. Normally the reactor was kept at a designated power level within certain limits. Occasionally the operators had to increase the power for a certain test. The shift supervisor and several different people would get together and discuss the change before boosting the power. All operators were required to maintain a Reactor Operator License from the Atomic Energy Commission. The license included six months of training, an eight-hour written exam, a four-hour walkaround, and testing on the reactor controls.

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 2.101 - Filing of application.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Reactors, the Director, Office of Nuclear Reactor Regulation, the Director, Office of Nuclear Material... Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and State... be requested to: (i) Submit to the Director, Office of Nuclear Reactor Regulation, Director, Office...

10 CFR 2.101 - Filing of application.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Reactors, the Director, Office of Nuclear Reactor Regulation, the Director, Office of Nuclear Material... Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and State... be requested to: (i) Submit to the Director, Office of Nuclear Reactor Regulation, Director, Office...

Weld monitor and failure detector for nuclear reactor system

DOEpatents

Sutton, Jr., Harry G.

1987-01-01

Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

Metabolic syndrome and quality of life: a systematic review.

PubMed

Saboya, Patrícia Pozas; Bodanese, Luiz Carlos; Zimmermann, Paulo Roberto; Gustavo, Andréia da Silva; Assumpção, Caroline Melo; Londero, Fernanda

2016-11-28

to present currently available evidence to verify the association between metabolic syndrome and quality of life. Cochrane Library, EMBASE, Medline and LILACS databases were studied for all studies investigating the association with metabolic syndrome and quality of life. Two blinded reviewers extracted data and one more was chosen in case of doubt. a total of 30 studies were included, considering inclusion and exclusion criteria, which involved 62.063 patients. Almost all studies suggested that metabolic syndrome is significantly associated with impaired quality of life. Some, however, found association only in women, or only if associated with depression or Body Mass Index. Merely one study did not find association after adjusted for confounding factors. although there are a few studies available about the relationship between metabolic syndrome and quality of life, a growing body of evidence has shown significant association between metabolic syndrome and the worsening of quality of life. However, it is necessary to carry out further longitudinal studies to confirm this association and verify whether this relationship is linear, or only an association factor. apresentar as evidências disponíveis atuais para verificar a associação entre síndrome metabólica e qualidade de vida. Cochrane Library, EMBASE, Medline e LILACS foram as bases de dados consultadas na identificação de todos os estudos que investigavam a associação entre síndrome metabólica e qualidade de vida. Dois revisores de forma independente e cegados extraíram os dados e, em caso de dúvidas, um outro revisor foi escolhido. um total de 30 estudos foram incluídos, considerando os critérios de inclusão e exclusão, os quais envolveram 62.063 pacientes. A maioria dos estudos sugerem que a síndrome metabólica é significativamente associada à piora da qualidade de vida. Alguns, no entanto, demonstram associação apenas em mulheres, ou somente se associadas à depressão ou índice de massa corporal. Entretanto, um estudo não demonstrou tal associação, após ajuste para os fatores confundidores. apesar de dispormos de poucos estudos a respeito da relação entre síndrome metabólica e qualidade de vida, um número crescente de evidências tem demonstrado uma significativa associação entre a síndrome metabólica e o prejuízo na qualidade de vida. Entretanto, é necessário que sejam conduzidos estudos longitudinais com objetivo de confirmar esta associação e, determinar se esta relação é linear ou somente um fator de associação. presentar las evidencias actualmente disponibles para verificar la asociación entre el síndrome metabólico y la calidad de vida. se consultaron las bases de datos Cochrane Library, EMBASE, Medline y LILACS para la identificación de todos los estudios que investigaban la asociación entre síndrome metabólico y calidad de vida. Dos revisores de tipo ciego extrajeron los datos y en caso de dudas se eligió un revisor adicional. un total de 30 estudios fueron incluidos, llevando en consideración los criterios de inclusión y exclusión, totalizando 62.063 pacientes. La mayoría de los estudios sugieren que el síndrome metabólico esta significativamente asociado al empeoramiento de la calidad de vida. Algunos, sin embargo muestran asociación solamente en mujeres, o sólo asociados a depresión o índice de masa corporal. También un estudio no mostró esta asociación después de ajustar por los factores confundentes. a pesar de tener a disposición pocos estudios en referencia a la relación entre síndrome metabólico y calidad de vida, un número creciente de evidencias ha demostrado una asociación significativa entre síndrome metabólico y perjuicio de la calidad de vida. Sin embargo es necesario que se realicen estudios longitudinales con el objetivo de confirmar esta asociación y determinar si esta relación es linear o solamente un factor de asociación.

Demonstration of Robustness and Integrated Operation of a Series-Bosch System

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, Matthew J.; Stanley, Christine; Barnett, Bill; Junaedi, Christian; Vilekar, Saurabh A.; Ryan, Kent

2016-01-01

Manned missions beyond low Earth orbit will require highly robust, reliable, and maintainable life support systems that maximize recycling of water and oxygen. Bosch technology is one option to maximize oxygen recovery, in the form of water, from metabolically-produced carbon dioxide (CO2). A two stage approach to Bosch, called Series-Bosch, reduces metabolic CO2 with hydrogen (H2) to produce water and solid carbon using two reactors: a Reverse Water-Gas Shift (RWGS) reactor and a carbon formation (CF) reactor. Previous development efforts demonstrated the stand-alone performance of a NASA-designed RWGS reactor designed for robustness against carbon formation, two membrane separators intended to maximize single pass conversion of reactants, and a batch CF reactor with both transit and surface catalysts. In the past year, Precision Combustion, Inc. (PCI) developed and delivered a RWGS reactor for testing at NASA. The reactor design was based on their patented Microlith® technology and was first evaluated under a Phase I Small Business Innovative Research (SBIR) effort in 2010. The RWGS reactor was recently evaluated at NASA to compare its performance and operating conditions with NASA's RWGS reactor. The test results will be provided in this paper. Separately, in 2015, a semi-continuous CF reactor was designed and fabricated at NASA based on the results from batch CF reactor testing. The batch CF reactor and the semi-continuous CF reactor were individually integrated with an upstream RWGS reactor to demonstrate the system operation and to evaluate performance. Here, we compare the performance and robustness to carbon formation of both RWGS reactors. We report the results of the integrated operation of a Series-Bosch system and we discuss the technology readiness level.

Reactor water cleanup system

DOEpatents

Gluntz, Douglas M.; Taft, William E.

1994-01-01

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

The role of nuclear reactors in space exploration and development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lipinski, R.J.

2000-07-01

The United States has launched more than 20 radioisotopic thermoelectric generators (RTGs) into space over the past 30 yr but has launched only one nuclear reactor, and that was in 1965. Russia has launched more than 30 reactors. The RTGs use the heat of alpha decay of {sup 238}Pu for power and typically generate <1 kW of electricity. Apollo, Pioneer, Voyager, Viking, Galileo, Ulysses, and Cassini all used RTGs. Space reactors use the fission energy of {sup 235}U; typical designs are for 100 to 1000 kW of electricity. The only US space reactor launch (SNAP-10A) was a demonstration mission. Onemore » reason for the lack of space reactor use by the United States was the lack of space missions that required high power. But, another was the assumed negative publicity that would accompany a reactor launch. The net result is that all space reactor programs after 1970 were terminated before an operating space reactor could be developed, and they are now many years from recovering the ability to build them. Two major near-term needs for space reactors are the human exploration of Mars and advanced missions to and beyond the orbit of Jupiter. To help obtain public acceptance of space reactors, one must correct some of the misconceptions concerning space reactors and convey the following facts to the public and to decision makers: Space reactors are 1000 times smaller in power and size than a commercial power reactor. A space reactor at launch is only as radioactive as a pile of dirt 60 m (200 ft) across. A space reactor contains no plutonium at launch. It does not become significantly radioactive until it is turned on, and it will be engineered so that no launch accident can turn it on, even if that means fueling it after launch. The reactor will not be turned on until it is in a high stable orbit or even on an earth-escape trajectory for some missions. The benefits of space reactors are that they give humanity a stairway to the planets and perhaps the stars. They open a new frontier for their children and their grandchildren. They pave the way for all life on earth to move out into the solar system. At one time, humans built and flew space reactors; it is time to do so again.« less

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

NASA Astrophysics Data System (ADS)

Shen, Yongjun; Lei, Lecheng; Zhang, Xingwang; Ding, Jiandong

2014-11-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10-9 mol/L and 0.61 × 10-9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10-2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10-2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation pathways were elucidated in light of the analyzed degradation products.

10 CFR 72.1 - Purpose.

Code of Federal Regulations, 2011 CFR

2011-01-01

... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.1 Purpose. The... receive, transfer, and possess power reactor spent fuel, power reactor-related Greater than Class C (GTCC... reactor spent fuel, high-level radioactive waste, power reactor-related GTCC waste, and other radioactive...

10 CFR 72.1 - Purpose.

Code of Federal Regulations, 2010 CFR

2010-01-01

... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.1 Purpose. The... receive, transfer, and possess power reactor spent fuel, power reactor-related Greater than Class C (GTCC... reactor spent fuel, high-level radioactive waste, power reactor-related GTCC waste, and other radioactive...

Unmixed fuel processors and methods for using the same

DOEpatents

Kulkarni, Parag Prakash; Cui, Zhe

2010-08-24

Disclosed herein are unmixed fuel processors and methods for using the same. In one embodiment, an unmixed fuel processor comprises: an oxidation reactor comprising an oxidation portion and a gasifier, a CO.sub.2 acceptor reactor, and a regeneration reactor. The oxidation portion comprises an air inlet, effluent outlet, and an oxygen transfer material. The gasifier comprises a solid hydrocarbon fuel inlet, a solids outlet, and a syngas outlet. The CO.sub.2 acceptor reactor comprises a water inlet, a hydrogen outlet, and a CO.sub.2 sorbent, and is configured to receive syngas from the gasifier. The regeneration reactor comprises a water inlet and a CO.sub.2 stream outlet. The regeneration reactor is configured to receive spent CO.sub.2 adsorption material from the gasification reactor and to return regenerated CO.sub.2 adsorption material to the gasification reactor, and configured to receive oxidized oxygen transfer material from the oxidation reactor and to return reduced oxygen transfer material to the oxidation reactor.

Thermionic switched self-actuating reactor shutdown system

DOEpatents

Barrus, Donald M.; Shires, Charles D.; Brummond, William A.

1989-01-01

A self-actuating reactor shutdown system incorporating a thermionic switched electromagnetic latch arrangement which is responsive to reactor neutron flux changes and to reactor coolant temperature changes. The system is self-actuating in that the sensing thermionic device acts directly to release (scram) the control rod (absorber) without reference or signal from the main reactor plant protective and control systems. To be responsive to both temperature and neutron flux effects, two detectors are used, one responsive to reactor coolant temperatures, and the other responsive to reactor neutron flux increase. The detectors are incorporated into a thermionic diode connected electrically with an electromagnetic mechanism which under normal reactor operating conditions holds the the control rod in its ready position (exterior of the reactor core). Upon reaching either a specified temperature or neutron flux, the thermionic diode functions to short-circuit the electromagnetic mechanism causing same to lose its holding power and release the control rod, which drops into the reactor core region under gravitational force.

Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

DOEpatents

Schreiber, R.B.; Fero, A.H.; Sejvar, J.

1997-12-16

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

DOEpatents

Schreiber, Roger B.; Fero, Arnold H.; Sejvar, James

1997-01-01

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soldevilla, M.; Salmons, S.; Espinosa, B.

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a uniquemore » repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)« less

Propellant actuated nuclear reactor steam depressurization valve

DOEpatents

Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

1992-01-01

A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

NEUTRONIC REACTOR CONSTRUCTION AND OPERATION

DOEpatents

West, J.M.; Weills, J.T.

1960-03-15

A method is given for operating a nuclear reactor having a negative coefficient of reactivity to compensate for the change in reactor reactivity due to the burn-up of the xenon peak following start-up of the reactor. When it is desired to start up the reactor within less than 72 hours after shutdown, the temperature of the reactor is lowered prior to start-up, and then gradually raised after start-up.

Thermal-hydraulic interfacing code modules for CANDU reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, W.S.; Gold, M.; Sills, H.

1997-07-01

The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

Ajustement statistique des simulations climatiques : l'exemple des précipitations saisonnières de l'Amérique tropicaleStatistical adjustment of simulated climate: example of seasonal rainfall of tropical America.

NASA Astrophysics Data System (ADS)

Moron, Vincent; Navarra, Antonio

2000-05-01

This study presents the skill of the seasonal rainfall of tropical America from an ensemble of three 34-year general circulation model (ECHAM 4) simulations forced with observed sea surface temperature between 1961 and 1994. The skill gives a first idea of the amount of potential predictability if the sea surface temperatures are perfectly known some time in advance. We use statistical post-processing based on the leading modes (extracted from Singular Value Decomposition of the covariance matrix between observed and simulated rainfall fields) to improve the raw skill obtained by simple comparison between observations and simulations. It is shown that 36-55 % of the observed seasonal variability is explained by the simulations on a regional basis. Skill is greatest for Brazilian Nordeste (March-May), but also for northern South America or the Caribbean basin in June-September or northern Amazonia in September-November for example.

97. ARAIII. ML1 reactor has been moved into GCRE reactor ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

97. ARA-III. ML-1 reactor has been moved into GCRE reactor building (ARA-608) for examination of corrosion on its underside and repair. May 24, 1963. Ineel photo no. 63-3485. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

NEUTRONIC REACTOR MANIPULATING DEVICE

DOEpatents

Ohlinger, L.A.

1962-08-01

A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

40 CFR 63.1406 - Reactor batch process vent provisions.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 40 Protection of Environment 11 2011-07-01 2011-07-01 false Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor... reactor batch process vent located at a new affected source shall control organic HAP emissions by...

78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-09

... Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft NUREG; request for comment. SUMMARY: The U.S..., Revision 10, ``Operator Licensing Examination Standards for Power Reactors.'' DATES: Submit comments [email protected] . Both of the Office of New Reactors; or Timothy Kolb, Office of Nuclear Reactor Regulation, U...

76 FR 55718 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-08

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels The ACRS Subcommittee on Materials, Metallurgy & Reactor...'' for reactor coolant system (RCS) components, as mentioned in 10 CFR 50 Appendix A, GDC-4. The...

40 CFR 63.1406 - Reactor batch process vent provisions.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor... reactor batch process vent located at a new affected source shall control organic HAP emissions by...

75 FR 58449 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-24

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels The ACRS Subcommittee on Materials, Metallurgy & Reactor... would result in a major inconvenience. Dated: September 17, 2010. Antonio Dias, Chief, Reactor Safety...

151. ARAIII Reactor building (ARA608) Details of reactor pit and ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

151. ARA-III Reactor building (ARA-608) Details of reactor pit and instrument plan. Aerojet-general 880-area/GCRE-608-T-19. Date: November 1958. Ineel index code no. 063-0608-25-013-102678. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

10 CFR 72.120 - General considerations.

Code of Federal Regulations, 2010 CFR

2010-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design... reactor-related GTCC waste in an ISFSI or to store spent fuel, high-level radioactive waste, or reactor... be designed to store spent fuel and/or solid reactor-related GTCC waste. (1) Reactor-related GTCC...

ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA660, INTERIOR. REACTOR INSIDE TANK. METAL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA-660, INTERIOR. REACTOR INSIDE TANK. METAL WORK PLATFORM ABOVE. THE REACTOR WAS IN A SMALL WATER-FILLED POOL. INL NEGATIVE NO. 66-6373. Unknown Photographer, ca. 1966 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

A Review of Gas-Cooled Reactor Concepts for SDI Applications

DTIC Science & Technology

1989-08-01

710 program .) Wire- Core Reactor (proposed by Rockwell). The wire- core reactor utilizes thin fuel wires woven between spacer wires to form an open...reactor is based on results of developmental studies of nuclear rocket propulsion systems. The reactor core is made up of annular fuel assemblies of...XE Addendum to Volume II. NERVA Fuel Development , Westinghouse Astronuclear Laboratory, TNR-230, July 15’ 1972. J I8- Rover Program Reactor Tests

Computer study of emergency shutdowns of a 60-kilowatt reactor Brayton space power system

NASA Technical Reports Server (NTRS)

Tew, R. C.; Jefferies, K. S.

1974-01-01

A digital computer study of emergency shutdowns of a 60-kWe reactor Brayton power system was conducted. Malfunctions considered were (1) loss of reactor coolant flow, (2) loss of Brayton system gas flow, (3)turbine overspeed, and (4) a reactivity insertion error. Loss of reactor coolant flow was the most serious malfunction for the reactor. Methods for moderating the reactor transients due to this malfunction are considered.

Function of university reactors in operator licensing training for nuclear utilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wicks, F.

1985-11-01

The director of the Division of the US Nuclear Regulatory Commission in generic letter 84-10, dated April 26, 1984, spoke the requirement that applicants for senior reactor operator licenses for power reactors shall have performed then reactor startups. Simulator startups were not acknowledged. Startups performed on a university reactor are acceptable. The content and results of a five-day program combining instruction and experiments with the Rensselaer reactor are summarized.

Numerical Simulations of a 96-rod Polysilicon CVD Reactor

NASA Astrophysics Data System (ADS)

Guoqiang, Tang; Cong, Chen; Yifang, Cai; Bing, Zong; Yanguo, Cai; Tihu, Wang

2018-05-01

With the rapid development of the photovoltaic industry, pressurized Siemens belljar-type polysilicon CVD reactors have been enlarged from 24 rods to 96 rods in less than 10 years aimed at much greater single-reactor productivity. A CFD model of an industry-scale 96-rod CVD reactor was established to study the internal temperature distribution and the flow field of the reactor. Numerical simulations were carried out and compared with actual growth results from a real CVD reactor. Factors affecting polysilicon depositions such as inlet gas injections, flow field, and temperature distribution in the CVD reactor are studied.

Reactor water cleanup system

DOEpatents

Gluntz, D.M.; Taft, W.E.

1994-12-20

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

Characteristics and Dose Levels for Spent Reactor Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coates, Cameron W

2007-01-01

Current guidance considers highly radioactive special nuclear materials to be those materials that, unshielded, emit a radiation dose [rate] measured at 1 m which exceeds 100 rem/h. Smaller, less massive fuel assemblies from research reactors can present a challenge from the point of view of self protection because of their size (lower dose, easier to handle) and the desirability of higher enrichments; however, a follow-on study to cross-compare dose trends of research reactors and power reactors was deemed useful to confirm/verify these trends. This paper summarizes the characteristics and dose levels of spent reactor fuels for both research reactors andmore » power reactors and extends previous studies aimed at quantifying expected dose rates from research reactor fuels worldwide.« less

Comparing the new generation accelerator driven subcritical reactor system (ADS) to traditional critical reactors

NASA Astrophysics Data System (ADS)

Kemah, Elif; Akkaya, Recep; Tokgöz, Seyit Rıza

2017-02-01

In recent years, the accelerator driven subcritical reactors have taken great interest worldwide. The Accelerator Driven System (ADS) has been used to produce neutron in subcritical state by the external proton beam source. These reactors, which are hybrid systems, are important in production of clean and safe energy and conversion of radioactive waste. The ADS with the selection of reliability and robust target materials have been the new generation of fission reactors. In addition, in the ADS Reactors the problems of long-lived radioactive fission products and waste actinides encountered in the fission process of the reactor during incineration can be solved, and ADS has come to the forefront of thorium as fuel for the reactors.

Reactor operation environmental information document

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haselow, J.S.; Price, V.; Stephenson, D.E.

1989-12-01

The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimalmore » impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.« less

Auxiliary reactor for a hydrocarbon reforming system

DOEpatents

Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Bentley, Jeffrey M.; Davis, Robert; Rumsey, Jennifer W.

2006-01-17

An auxiliary reactor for use with a reformer reactor having at least one reaction zone, and including a burner for burning fuel and creating a heated auxiliary reactor gas stream, and heat exchanger for transferring heat from auxiliary reactor gas stream and heat transfer medium, preferably two-phase water, to reformer reaction zone. Auxiliary reactor may include first cylindrical wall defining a chamber for burning fuel and creating a heated auxiliary reactor gas stream, the chamber having an inlet end, an outlet end, a second cylindrical wall surrounding first wall and a second annular chamber there between. The reactor being configured so heated auxiliary reactor gas flows out the outlet end and into and through second annular chamber and conduit which is disposed in second annular chamber, the conduit adapted to carry heat transfer medium and being connectable to reformer reaction zone for additional heat exchange.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lister, Tedd E; Parkman, Jacob A; Diaz Aldana, Luis A

A method of recovering metals from electronic waste comprises providing a powder comprising electronic waste in at least a first reactor and a second reactor and providing an electrolyte comprising at least ferric ions in an electrochemical cell in fluid communication with the first reactor and the second reactor. The method further includes contacting the powders within the first reactor and the second reactor with the electrolyte to dissolve at least one base metal from each reactor into the electrolyte and reduce at least some of the ferric ions to ferrous ions. The ferrous ions are oxidized at an anodemore » of the electrochemical cell to regenerate the ferric ions. The powder within the second reactor comprises a higher weight percent of the at least one base metal than the powder in the first reactor. Additional methods of recovering metals from electronic waste are also described, as well as an apparatus of recovering metals from electronic waste.« less

High-intensity power-resolved radiation imaging of an operational nuclear reactor.

PubMed

Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

2015-10-09

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

Exploratory study of several advanced nuclear-MHD power plant systems.

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.; Rosa, R. J.; Yang, Y. Y.

1973-01-01

In order for efficient multimegawatt closed cycle nuclear-MHD systems to become practical, long-life gas cooled reactors with exit temperatures of about 2500 K or higher must be developed. Four types of nuclear reactors which have the potential of achieving this goal are the NERVA-type solid core reactor, the colloid core (rotating fluidized bed) reactor, the 'light bulb' gas core reactor, and the 'coaxial flow' gas core reactor. Research programs aimed at developing these reactors have progressed rapidly in recent years so that prototype power reactors could be operating by 1980. Three types of power plant systems which use these reactors have been analyzed to determine the operating characteristics, critical parameters and performance of these power plants. Overall thermal efficiencies as high as 80% are projected, using an MHD turbine-compressor cycle with steam bottoming, and slightly lower efficiencies are projected for an MHD motor-compressor cycle.

High-intensity power-resolved radiation imaging of an operational nuclear reactor

PubMed Central

Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

2015-01-01

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

NASA Astrophysics Data System (ADS)

Kim, Kwangmin; Go, Byeong-Soo; Sung, Hae-Jin; Park, Hea-chul; Kim, Seokho; Lee, Sangjin; Jin, Yoon-Su; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2014-09-01

This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

A comparison of the technological effectiveness of dairy wastewater treatment in anaerobic UASB reactor and anaerobic reactor with an innovative design.

PubMed

Jedrzejewska-Cicinska, M; Kozak, K; Krzemieniewski, M

2007-10-01

The present research was an investigation of the influence of an innovative design of reactor filled with polyethylene (PE) granulate on model dairy wastewater treatment efficiency under anaerobic conditions compared to that obtained in a typical UASB reactor. The experiment was conducted at laboratory scale. An innovative reactor was designed with the reaction chamber inclined 30 degrees in relation to the ground with upward waste flow and was filled with PE granular material. Raw model dairy wastewater was fed to two anaerobic reactors of different design at the organic loading rate of 4 kg COD m(-3)d(-1). Throughout the experiment, a higher removal efficiency of organic compounds was observed in the reactor with an innovative design and it was higher by 7.1% on average than in the UASB reactor. The total suspended solids was lower in the wastewater treated in the anaerobic reactor with the innovative design. Applying a PE granulated filling in the chamber of the innovative reactor contributed to an even distribution of sludge biomass in the reactor, reducing washout of anaerobic sludge biomass from the reaction chamber and giving a higher organic compounds removal efficiency.

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

NASA Astrophysics Data System (ADS)

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.

Development of toroid-type HTS DC reactor series for HVDC system

NASA Astrophysics Data System (ADS)

Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2015-11-01

This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

Experiences in utilization of research reactors in Yugoslavia

DOE Office of Scientific and Technical Information (OSTI.GOV)

Copic, M.; Gabrovsek, Z.; Pop-Jordanov, J.

1971-06-15

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied bymore » means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro envisage the role of research reactors in the promotion of nuclear power programs in relation to the size of the program, the competence of domestic industries and the degree of independence where fuel supply is concerned. (author)« less

Association between diabetes and tuberculosis: case-control study.

PubMed

Pereira, Susan Martins; Araújo, Gleide Santos de; Santos, Carlos Antônio de Souza Teles; Oliveira, Maeli Gomes de; Barreto, Maurício Lima

2016-12-22

To test the association between diabetes and tuberculosis. It is a case-control study, matched by age and sex. We included 323 new cases of tuberculosis with positive results for bacilloscopy. The controls were 323 respiratory symptomatic patients with negative bacilloscopy, from the same health services, such as: ambulatory cases from three referral hospitals and six basic health units responsible for the notifications of new cases of tuberculosis in Salvador, Bahia. Data collection occurred between 2008 and 2010. The instruments used were structured interview, including clinical data, capillary blood glucose (during fasting or postprandial), and the CAGE questionnaire for screening of abusive consumption of alcohol. Descriptive, exploratory, and multivariate analysis was performed using conditional logistic regression. The average age of the cases was 38.5 (SD = 14.2) years and of the controls, 38.5 (SD = 14.3) years. Among cases and controls, most subjects (61%) were male. In univariate analysis we found association between the occurrence of diabetes and tuberculosis (OR = 2.37; 95%CI 1.04-5.42), which remained statistically significant after adjustment for potential confounders (OR = 3.12; 95%CI 1.12-7.94). The association between diabetes and tuberculosis can hinder the control of tuberculosis, contributing to the maintainance of the disease burden. The situation demands increasing early detection of diabetes among people with tuberculosis, in an attempt to improve disease control strategies. Testar a associação entre diabetes e tuberculose. Trata-se de estudo caso-controle, pareado por idade e sexo. Foram incluídos 323 casos novos de tuberculose com resultados positivos à baciloscopia. Os controles foram 323 sintomáticos respiratórios com baciloscopia negativa, oriundos dos mesmos serviços de saúde dos casos: ambulatórios de três hospitais de referência e seis unidades básicas de saúde responsáveis pelas notificações dos casos novos de tuberculose em Salvador, Bahia. A coleta de dados ocorreu entre 2008 e 2010. Os instrumentos utilizados foram entrevista estruturada, incluindo dados clínicos, glicemia capilar (em jejum ou pós-prandial) e o questionário CAGE para triagem de consumo abusivo de álcool. Foi realizada análise descritiva, exploratória e multivariada utilizando-se de regressão logística condicional. A média de idade dos casos foi 38,5 (DP = 14,2) anos e dos controles, 38,5 (DP = 14,3) anos. Tanto entre os casos quanto entre os controles, a maioria (61%) dos indivíduos era do sexo masculino. Na análise univariada, houve associação entre ocorrência de diabetes e de tuberculose (OR = 2,37; IC95% 1,04-5,42), que permaneceu estatisticamente significante após ajuste pelos potenciais confundidores (OR = 3,12; IC95% 1,12-7,94). A associação entre diabetes e tuberculose pode dificultar o controle da tuberculose, contribuindo para manutenção da elevada carga da doença. A situação demanda intensificação da detecção precoce de diabetes entre pessoas com tuberculose, na tentativa de maior efetividade das estratégias de controle da doença.

Increased ultrasensitive C-reactive protein is not associated with obesity in hospitalized heart failure patients.

PubMed

Schommer, Vânia Ames; Stein, Airton Tetelbom; Marcadenti, Aline; Wittke, Estefania Inez; Galvão, André Luís Câmara; Rosito, Guido Bernardo Aranha

2016-01-01

To evaluate the association between obesity and levels of high-sensitivity C-reactive protein (hs-CRP) in patients with heart failure admitted to a tertiary hospital. Cross-sectional study with a consecutive sampling of hospitalized patients with heart failure. Sociodemographic and clinical data were collected, and the nutritional status was assessed through indicators such as body mass index (in kg/m2), waist circumference (in cm), waist-hip ratio, triceps skinfold (in mm) and subscapularis skinfold (in mm). Neck circumference (in cm) was measured as well as serum levels of hs-CRP, in mg/L. Among 123 patients, the mean age was 61.9±12.3 years and 60.2% were male. The median of hs-CRP was 8.87mg/L (3.34 to 20.01). A tendency to an inverse correlation between neck circumference and hs-CRP was detected (r=-0.167; p=0.069). In the multiple linear regression analysis, after adjustment for age, disease severity (NYHA classification III and IV, low ejection fraction, left ventricular dysfunction during diastole), and infectious conditions there was an inverse association between hs-CRP and neck circumference (ß=-0.196; p=0.03) and subscapularis skinfold (ß=-0.005; p=0.01) in the total sample, which was not maintained after the stratification by sex. Increased levels of hs-CRP in patients hospitalized for heart failure were not associated with obesity. Avaliar a associação entre obesidade e níveis de proteína c-reativa ultrassensível (PCR-us) em pacientes com insuficiência cardiac admitidos em um hospital terciário. Estudo transversal com amostragem consecutiva de pacientes com insuficiência cardíaca hospitalizados. Foram coletados dados sociodemográficos e clínicos, e o estado nutricional foi avaliado por meio de indicadores como índice de massa corporal (em kg/m2), circunferência da cintura (em cm), razão cintura-quadril, dobra cutânea tricipital (em mm) e dobra cutânea subescapular (em mm). Circunferência do pescoço (em cm) foi aferida bem como níveis séricos de PCR-us, em mg/L. Em 123 pacientes, a média da idade foi 61,9±12,3 anos, e 60,2% eram do sexo masculino. A mediana de PCR-us foi de 8,87mg/L (3,34 a 20,01). Detectou-se tendência à correlação inversa entre circunferência do pescoço e PCR-us (r=-0,167; p=0,069). Na análise por regressão linear múltipla, após ajustes para idade, gravidade da doença (classificação NYHA III e IV, fração de ejeção baixa, disfunção ventricular esquerda durante a diástole) e quadros infecciosos, houve associação inversa entre PCR-us e circunferência do pescoço (ß=-0,196; p=0,03) e dobra cutânea subescapular (ß=-0,005; p=0,01) na amostra total, que não se manteve após estratificação para sexo. O aumento dos níveis de PCR-us em pacientes hospitalizados por insuficiência cardíaca não se associou à obesidade.

Cross-Cultural adaptation of the General Functioning Scale of the Family.

PubMed

Pires, Thiago; Assis, Simone Gonçalves de; Avanci, Joviana Quintes; Pesce, Renata Pires

2016-06-27

To describe the process of cross-cultural adaptation of the General Functioning Scale of the Family, a subscale of the McMaster Family Assessment Device, for the Brazilian population. The General Functioning Scale of the Family was translated into Portuguese and administered to 500 guardians of children in the second grade of elementary school in public schools of Sao Gonçalo, Rio de Janeiro, Southeastern Brazil. The types of equivalences investigated were: conceptual and of items, semantic, operational, and measurement. The study involved discussions with experts, translations and back-translations of the instrument, and psychometric assessment. Reliability and validity studies were carried out by internal consistency testing (Cronbach's alpha), Guttman split-half correlation model, Pearson correlation coefficient, and confirmatory factor analysis. Associations between General Functioning of the Family and variables theoretically associated with the theme (father's or mother's drunkenness and violence between parents) were estimated by odds ratio. Semantic equivalence was between 90.0% and 100%. Cronbach's alpha ranged from 0.79 to 0.81, indicating good internal consistency of the instrument. Pearson correlation coefficient ranged between 0.303 and 0.549. Statistical association was found between the general functioning of the family score and the theoretically related variables, as well as good fit quality of the confirmatory analysis model. The results indicate the feasibility of administering the instrument to the Brazilian population, as it is easy to understand and a good measurement of the construct of interest. Descrever o processo de adaptação transcultural da escala de Funcionamento Geral da Família, subescala da McMaster Family Assessment Device, para a população brasileira. A escala de Funcionamento Geral da Família, original no idioma inglês, foi traduzida para o português e aplicada a 500 responsáveis de crianças do segundo ano do ensino fundamental de escolas públicas do município de São Gonçalo, Rio de Janeiro. Os tipos de equivalências investigados foram: conceitual e de itens, semântica, operacional, e mensuração. O estudo envolveu discussão com especialistas, traduções e retraduções do instrumento e avaliação psicométrica. Foi realizado estudo de confiabilidade e validade, por testagem da consistência interna (alpha de Cronbach), da correlação pelo método de split-half de Guttman e do coeficiente de correlação de Pearson, e por análise fatorial confirmatória. Associações entre Funcionamento Geral da Família e variáveis teoricamente associadas ao tema (embriaguez do pai ou da mãe e violência entre os pais) foram analisadas por estimação de razão de chance. A equivalência semântica encontrada foi entre 90,0% e 100%. O alfa de Cronbach variou de 0,79 a 0,81, indicando boa consistência interna do instrumento. O coeficiente de correlação de Pearson oscilou entre 0,303 e 0,549. Encontrou-se associação estatística entre o escore de funcionamento geral da família e as variáveis teoricamente relacionadas e constatou-se boa qualidade dos ajustes do modelo de análise confirmatória. Os resultados obtidos indicam a aplicabilidade do instrumento na população brasileira, que se mostra de fácil compreensão e boa aferição do constructo de interesse.

Nuclear Reactor Physics

NASA Astrophysics Data System (ADS)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

SELF-REGULATING BOILING-WATER NUCLEAR REACTORS

DOEpatents

Ransohoff, J.A.; Plawchan, J.D.

1960-08-16

A boiling-water reactor was designed which comprises a pressure vessel containing a mass of water, a reactor core submerged within the water, a reflector tank disposed within the reactor, the reflector tank being open at the top to the interior of the pressure vessel, and a surge tank connected to the reflector tank. In operation the reflector level changes as a function of the pressure witoin the reactor so that the reactivity of the reactor is automatically controlled.

REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT

DOEpatents

Loeb, E.

1961-01-17

A power generator in the form of a flywheel with four reactors positioned about its rim is described. The reactors are so positioned that steam, produced in the reactor, exists tangentially to the flywheel, giving it a rotation. The reactors are incompletely moderated without water. The water enters the flywheel at its axis, under sufficient pressure to force it through the reactors, where it is converted to steam. The fuel consists of parallel twisted ribbons assembled to approximate a cylinder.

System and method for air temperature control in an oxygen transport membrane based reactor

DOEpatents

Kelly, Sean M

2016-09-27

A system and method for air temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

System and method for temperature control in an oxygen transport membrane based reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly, Sean M.

A system and method for temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

Navy Nuclear-Powered Surface Ships: Background, Issues, and Options for Congress

DTIC Science & Technology

2010-09-29

to design a smaller scale version of a naval pressurized water reactor , or to design a new reactor type potentially using a thorium liquid salt...integrated nuclear power system capable of use on destroyer- sized vessels either using a pressurized water reactor or a thorium liquid salt reactor ...nuclear reactors for Navy surface ships. The text of Section 246 is as follows: SEC. 246. STUDY ON THORIUM -LIQUID FUELED REACTORS FOR NAVAL FORCES

Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1995-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduate use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The low power output of the reactor and an ergonomicmore » facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor`s Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced.« less

Modification of UASB reactor by using CFD simulations for enhanced treatment of municipal sewage.

PubMed

Das, Suprotim; Sarkar, Supriya; Chaudhari, Sanjeev

2018-02-01

Up-flow anaerobic sludge blanket (UASB) has been in use since last few decades for the treatment of organic wastewaters. However, the performance of UASB reactor is quite low for treatment of low strength wastewaters (LSWs) due to less biogas production leading to poor mixing. In the present research work, a modification was done in the design of UASB to improve mixing of reactor liquid which is important to enhance the reactor performance. The modified UASB (MUASB) reactor was designed by providing a slanted baffle along the height of the reactor having an angle of 5.7° with the vertical wall. A two-dimensional computational fluid dynamics (CFD) simulation of three phase gas-liquid-solid flow in MUASB reactor was performed and compared with conventional UASB reactor. The CFD study indicated better mixing in terms of vorticity magnitude in MUASB reactor as compared to conventional UASB, which was reflected in the reactor performance. The performance of MUASB was compared with conventional UASB reactor for the onsite treatment of domestic sewage as LSW. Around 16% higher total chemical oxygen demand removal efficiency was observed in MUASB reactor as compared to conventional UASB during this study. Therefore, this MUASB model demonstrates a qualitative relationship between mixing and performance during the treatment of LSW. From the study, it seems that MUASB holds promise for field applications.

10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

Code of Federal Regulations, 2010 CFR

2010-01-01

... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec...

10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

Code of Federal Regulations, 2011 CFR

2011-01-01

... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec...

78 FR 26811 - Dow Chemical Company, Dow TRIGA Research Reactor; License Renewal for the Dow Chemical TRIGA...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-08

... Research Reactor; License Renewal for the Dow Chemical TRIGA Research Reactor; Supplemental Information and... 20, 2012 (77 FR 42771), ``License Renewal for the Dow Chemical TRIGA Research Reactor,'' to inform... Chemical Company which would authorize continued operation of the Dow TRIGA Research Reactor. The notice...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Amount of financial protection required for other reactors... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Amount of financial protection required for other reactors... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000...

PBF Reactor Building (PER620). Camera faces north into highbay/reactor pit ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). Camera faces north into high-bay/reactor pit area. Inside from for reactor enclosure is in place. Photographer: John Capek. Date: March 15, 1967. INEEL negative no. 67-1769 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Reactor safety method

DOEpatents

Vachon, Lawrence J.

1980-03-11

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

155. ARAIII Reactor building (ARA608) Details of reactor pit showing ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

155. ARA-III Reactor building (ARA-608) Details of reactor pit showing tray supports and fuel element storage rack. Aerojet-general 880-area/GCRE-608-MS-2. Date: November 1958. Ineel index code no. 063-0608-40-013-102625. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

World Energy Data System (WENDS). Volume XI. Nuclear fission program summaries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1979-06-01

Brief management and technical summaries of nuclear fission power programs are presented for nineteen countries. The programs include the following: fuel supply, resource recovery, enrichment, fuel fabrication, light water reactors, heavy water reactors, gas cooled reactors, breeder reactors, research and test reactors, spent fuel processing, waste management, and safety and environment. (JWR)

Breeder Reactors, Understanding the Atom Series.

ERIC Educational Resources Information Center

Mitchell, Walter, III; Turner, Stanley E.

The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000... chapter to operate two or more nuclear reactors at the same location, the total financial protection...

10 CFR 2.103 - Action on applications for byproduct, source, special nuclear material, facility and operator...

Code of Federal Regulations, 2011 CFR

2011-01-01

..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Nuclear... of this chapter, see § 2.106(d). (b) If the Director, Office of Nuclear Reactor Regulation, Director...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000... chapter to operate two or more nuclear reactors at the same location, the total financial protection...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000... chapter to operate two or more nuclear reactors at the same location, the total financial protection...

10 CFR 2.103 - Action on applications for byproduct, source, special nuclear material, facility and operator...

Code of Federal Regulations, 2012 CFR

2012-01-01

..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Nuclear... of this chapter, see § 2.106(d). (b) If the Director, Office of Nuclear Reactor Regulation, Director...

Computer model of catalytic combustion/Stirling engine heater head

NASA Technical Reports Server (NTRS)

Chu, E. K.; Chang, R. L.; Tong, H.

1981-01-01

The basic Acurex HET code was modified to analyze specific problems for Stirling engine heater head applications. Specifically, the code can model: an adiabatic catalytic monolith reactor, an externally cooled catalytic cylindrical reactor/flat plate reactor, a coannular tube radiatively cooled reactor, and a monolithic reactor radiating to upstream and downstream heat exchangers.

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

MONTE CARLO SIMULATIONS OF PERIODIC PULSED REACTOR WITH MOVING GEOMETRY PARTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cao, Yan; Gohar, Yousry

2015-11-01

In a periodic pulsed reactor, the reactor state varies periodically from slightly subcritical to slightly prompt supercritical for producing periodic power pulses. Such periodic state change is accomplished by a periodic movement of specific reactor parts, such as control rods or reflector sections. The analysis of such reactor is difficult to perform with the current reactor physics computer programs. Based on past experience, the utilization of the point kinetics approximations gives considerable errors in predicting the magnitude and the shape of the power pulse if the reactor has significantly different neutron life times in different zones. To accurately simulate themore » dynamics of this type of reactor, a Monte Carlo procedure using the transfer function TRCL/TR of the MCNP/MCNPX computer programs is utilized to model the movable reactor parts. In this paper, two algorithms simulating the geometry part movements during a neutron history tracking have been developed. Several test cases have been developed to evaluate these procedures. The numerical test cases have shown that the developed algorithms can be utilized to simulate the reactor dynamics with movable geometry parts.« less

Pressurized fluidized bed reactor

DOEpatents

Isaksson, J.

1996-03-19

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

Pressurized fluidized bed reactor

DOEpatents

Isaksson, Juhani

1996-01-01

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

Reactor vibration reduction based on giant magnetostrictive materials

NASA Astrophysics Data System (ADS)

Rongge, Yan; Weiying, Liu; Yuechao, Wu; Menghua, Duan; Xiaohong, Zhang; Lihua, Zhu; Ling, Weng; Ying, Sun

2017-05-01

The vibration of reactors not only produces noise pollution, but also affects the safe operation of reactors. Giant magnetostrictive materials can generate huge expansion and shrinkage deformation in a magnetic field. With the principle of mutual offset between the giant magnetostrictive force produced by the giant magnetostrictive material and the original vibration force of the reactor, the vibration of the reactor can be reduced. In this paper, magnetization and magnetostriction characteristics in silicon steel and the giant magnetostrictive material are measured, respectively. According to the presented magneto-mechanical coupling model including the electromagnetic force and the magnetostrictive force, reactor vibration is calculated. By comparing the vibration of the reactor with different inserted materials in the air gaps between the reactor cores, the vibration reduction effectiveness of the giant magnetostrictive material is validated.

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Su'ud, Zaki; Anshari, Rio

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environmentmore » such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

NASA Astrophysics Data System (ADS)

Su'ud, Zaki; Anshari, Rio

2012-06-01

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wichman, K.; Tsao, J.; Mayfield, M.

The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRCmore » and the International Piping Integrity Research Group is also briefly summarized.« less

Solution of heat removal from nuclear reactors by natural convection

NASA Astrophysics Data System (ADS)

Zitek, Pavel; Valenta, Vaclav

2014-03-01

This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR).The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor) for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

Imaging Fukushima Daiichi reactors with muons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.

2013-05-15

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi tomore » make this determination in the near future.« less

Imaging Fukushima Daiichi reactors with muons

NASA Astrophysics Data System (ADS)

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Lukić, Zarija; Masuda, Koji; Milner, Edward C.; Morris, Christopher L.; Perry, John O.

2013-05-01

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fitzpatrick, F.C.; Gray, D.D.; Hyndman, J.R.

The thermal, ecological, and social impacts of a 40-reactor NEC are compared to impacts from four 10-reactor NECs and ten 4-reactor power plants. The comparison was made for surrogate sites in western Tennessee. The surrogate site for the 40-reactor NEC is located on Kentucky Lake. A layout is postulated for ten clusters of four reactors each with 2.5-mile spacing between clusters. The plants use natural-draft cooling towers. A transmission system is proposed for delivering the power (48,000 MW) to five load centers. Comparable transmission systems are proposed for the 10-reactor NECs and the 4-reactor dispersed sites delivering power to themore » same load centers. (auth)« less

Digital computer study of nuclear reactor thermal transients during startup of 60-kWe Brayton power conversion system

NASA Technical Reports Server (NTRS)

Jefferies, K. S.; Tew, R. C.

1974-01-01

A digital computer study was made of reactor thermal transients during startup of the Brayton power conversion loop of a 60-kWe reactor Brayton power system. A startup procedure requiring the least Brayton system complication was tried first; this procedure caused violations of design limits on key reactor variables. Several modifications of this procedure were then found which caused no design limit violations. These modifications involved: (1) using a slower rate of increase in gas flow; (2) increasing the initial reactor power level to make the reactor respond faster; and (3) appropriate reactor control drum manipulation during the startup transient.

Hybrid adsorptive membrane reactor

NASA Technical Reports Server (NTRS)

Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

2011-01-01

A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

Hybrid adsorptive membrane reactor

DOEpatents

Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

2011-03-01

A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

NASA Astrophysics Data System (ADS)

Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

Reactor performances and microbial communities of biogas reactors: effects of inoculum sources.

PubMed

Han, Sheng; Liu, Yafeng; Zhang, Shicheng; Luo, Gang

2016-01-01

Anaerobic digestion is a very complex process that is mediated by various microorganisms, and the understanding of the microbial community assembly and its corresponding function is critical in order to better control the anaerobic process. The present study investigated the effect of different inocula on the microbial community assembly in biogas reactors treating cellulose with various inocula, and three parallel biogas reactors with the same inoculum were also operated in order to reveal the reproducibility of both microbial communities and functions of the biogas reactors. The results showed that the biogas production, volatile fatty acid (VFA) concentrations, and pH were different for the biogas reactors with different inocula, and different steady-state microbial community patterns were also obtained in different biogas reactors as reflected by Bray-Curtis similarity matrices and taxonomic classification. It indicated that inoculum played an important role in shaping the microbial communities of biogas reactor in the present study, and the microbial community assembly in biogas reactor did not follow the niche-based ecology theory. Furthermore, it was found that the microbial communities and reactor performances of parallel biogas reactors with the same inoculum were different, which could be explained by the neutral-based ecology theory and stochastic factors should played important roles in the microbial community assembly in the biogas reactors. The Bray-Curtis similarity matrices analysis suggested that inoculum affected more on the microbial community assembly compared to stochastic factors, since the samples with different inocula had lower similarity (10-20 %) compared to the samples from the parallel biogas reactors (30 %).

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

Bhattacharyya, Samit K.; Lipinski, Walter C.; Hanan, Nelson A.

The present plant protection system (PPS) has been defined for use in the TREAT-upgrade (TU) reactor for controlled transient operation of reactor-fuel behavior testing under simulated reactor-accident conditions. A PPS with energy-dependent trip set points lowered worst-case clad temperatures by as much as 180 K, relative to the use of conventional fixed-level trip set points. The multilayered multilevel protection strategy represents the state-of-the-art in terrestrial transient reactor protection systems, and should be applicable to multi-MW space reactors.

Nuclear Energy Policy

DTIC Science & Technology

2009-12-10

Small Modular Reactors Rising cost estimates for large conventional nuclear power plants—widely projected to be $6 billion or more—have contributed to growing interest in proposals for smaller, modular reactors. Ranging from about 40 to 350 megawatts of electrical capacity, such reactors would be only a fraction of the size of current commercial reactors. Several modular reactors would be installed together to make up a power block with a single control room, under most concepts. Modular reactor concepts would use a variety of technologies,

Research and proposal on selective catalytic reduction reactor optimization for industrial boiler.

PubMed

Yang, Yiming; Li, Jian; He, Hong

2017-08-24

The advanced computational fluid dynamics (CFD) software STAR-CCM+ was used to simulate a denitrification (De-NOx) project for a boiler in this paper, and the simulation result was verified based on a physical model. Two selective catalytic reduction (SCR) reactors were developed: reactor 1 was optimized and reactor 2 was developed based on reactor 1. Various indicators, including gas flow field, ammonia concentration distribution, temperature distribution, gas incident angle, and system pressure drop were analyzed. The analysis indicated that reactor 2 was of outstanding performance and could simplify developing greatly. Ammonia injection grid (AIG), the core component of the reactor, was studied; three AIGs were developed and their performances were compared and analyzed. The result indicated that AIG 3 was of the best performance. The technical indicators were proposed for SCR reactor based on the study. Flow filed distribution, gas incident angle, and temperature distribution are subjected to SCR reactor shape to a great extent, and reactor 2 proposed in this paper was of outstanding performance; ammonia concentration distribution is subjected to ammonia injection grid (AIG) shape, and AIG 3 could meet the technical indicator of ammonia concentration without mounting ammonia mixer. The developments above on the reactor and the AIG are both of great application value and social efficiency.

REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nichols, T.; Beals, D.; Sternat, M.

2011-07-18

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.« less

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note—A nuclear reactor basically includes the items within or attached directly to the reactor vessel, the equipment which controls the...

78 FR 20959 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-08

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... Committee on Reactor Safeguards. [FR Doc. 2013-08131 Filed 4-5-13; 8:45 am] BILLING CODE 7590-01-P ...

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note—A nuclear reactor basically includes the items within or attached directly to the reactor vessel, the equipment which controls the...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental...

NEUTRONIC REACTOR

DOEpatents

Fermi, E.; Zinn, W.H.; Anderson, H.L.

1958-09-16

Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

76 FR 70331 - List of Approved Spent Fuel Storage Casks: MAGNASTOR ® System, Revision 2

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-14

... various boron-10 areal densities for use with Pressurized Water Reactor and Boiling Water Reactor baskets... add various boron-10 areal densities for use with Pressurized Water Reactor and Boiling Water Reactor....1.1 to add various boron-10 areal densities for use with Pressurized Water Reactor and Boiling Water...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note: A nuclear reactor... core of a nuclear reactor and capable of withstanding the operating pressure of the primary coolant. (2...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.

1958-04-22

A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

Fütterer, Michael A.; D'Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco; Raison, Philippe; Bakker, Klaas; de Groot, Sander; Klaassen, Frodo

2009-07-01

This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

Using thermal balance model to determine optimal reactor volume and insulation material needed in a laboratory-scale composting reactor.

PubMed

Wang, Yongjiang; Pang, Li; Liu, Xinyu; Wang, Yuansheng; Zhou, Kexun; Luo, Fei

2016-04-01

A comprehensive model of thermal balance and degradation kinetics was developed to determine the optimal reactor volume and insulation material. Biological heat production and five channels of heat loss were considered in the thermal balance model for a representative reactor. Degradation kinetics was developed to make the model applicable to different types of substrates. Simulation of the model showed that the internal energy accumulation of compost was the significant heat loss channel, following by heat loss through reactor wall, and latent heat of water evaporation. Lower proportion of heat loss occurred through the reactor wall when the reactor volume was larger. Insulating materials with low densities and low conductive coefficients were more desirable for building small reactor systems. Model developed could be used to determine the optimal reactor volume and insulation material needed before the fabrication of a lab-scale composting system. Copyright © 2016 Elsevier Ltd. All rights reserved.

Utilization of Stop-flow Micro-tubing Reactors for the Development of Organic Transformations.

PubMed

Toh, Ren Wei; Li, Jie Sheng; Wu, Jie

2018-01-04

A new reaction screening technology for organic synthesis was recently demonstrated by combining elements from both continuous micro-flow and conventional batch reactors, coined stop-flow micro-tubing (SFMT) reactors. In SFMT, chemical reactions that require high pressure can be screened in parallel through a safer and convenient way. Cross-contamination, which is a common problem in reaction screening for continuous flow reactors, is avoided in SFMT. Moreover, the commercially available light-permeable micro-tubing can be incorporated into SFMT, serving as an excellent choice for light-mediated reactions due to a more effective uniform light exposure, compared to batch reactors. Overall, the SFMT reactor system is similar to continuous flow reactors and more superior than batch reactors for reactions that incorporate gas reagents and/or require light-illumination, which enables a simple but highly efficient reaction screening system. Furthermore, any successfully developed reaction in the SFMT reactor system can be conveniently translated to continuous-flow synthesis for large scale production.

Employing ISRU Models to Improve Hardware Design

NASA Technical Reports Server (NTRS)

Linne, Diane L.

2010-01-01

An analytical model for hydrogen reduction of regolith was used to investigate the effects of several key variables on the energy and mass performance of reactors for a lunar in-situ resource utilization oxygen production plant. Reactor geometry, reaction time, number of reactors, heat recuperation, heat loss, and operating pressure were all studied to guide hardware designers who are developing future prototype reactors. The effects of heat recuperation where the incoming regolith is pre-heated by the hot spent regolith before transfer was also investigated for the first time. In general, longer reaction times per batch provide a lower overall energy, but also result in larger and heavier reactors. Three reactors with long heat-up times results in similar energy requirements as a two-reactor system with all other parameters the same. Three reactors with heat recuperation results in energy reductions of 20 to 40 percent compared to a three-reactor system with no heat recuperation. Increasing operating pressure can provide similar energy reductions as heat recuperation for the same reaction times.

NEUTRONIC REACTOR SYSTEM

DOEpatents

Treshow, M.

1959-02-10

A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belles, Randy; Jain, Prashant K.; Powers, Jeffrey J.

The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments usingmore » equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.« less

The IRIS Spool-Type Reactor Coolant Pump

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kujawski, J.M.; Kitch, D.M.; Conway, L.E.

2002-07-01

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safetymore » to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called 'spool' pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper. (authors)« less

10 CFR 171.3 - Scope.

Code of Federal Regulations, 2010 CFR

2010-01-01

... holding an operating license for a power reactor, test reactor or research reactor issued under part 50 of... authorizes operation of a power reactor. The regulations in this part also apply to any person holding a...

Advanced Test Reactor Tour

ScienceCinema

Miley, Don

2017-12-21

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored.

Looking Southwest at Reactor Box Furnaces With Reactor Boxes and ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

Looking Southwest at Reactor Box Furnaces With Reactor Boxes and Repossessed Uranium in Recycle Recovery Building - Hematite Fuel Fabrication Facility, Recycle Recovery Building, 3300 State Road P, Festus, Jefferson County, MO

Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

NASA Astrophysics Data System (ADS)

Tong, H.; Snow, G. C.; Chu, E. K.; Chang, R. L. S.; Angwin, M. J.; Pessagno, S. L.

1981-09-01

Durable catalytic reactors for advanced gas turbine engines were developed. Objectives were: to evaluate furnace aging as a cost effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1,000 hours of combustion durability, and define a catalytic reactor system with a high probability of successful integration into an automotive gas turbine engine. Fourteen different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel for combustion conditions. Eight additional catalytic reactors were evaluated and one of these was successfully combustion tested on propane fuel. This durability reactor used graded cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

Moving bed reactor setup to study complex gas-solid reactions.

PubMed

Gupta, Puneet; Velazquez-Vargas, Luis G; Valentine, Charles; Fan, Liang-Shih

2007-08-01

A moving bed scale reactor setup for studying complex gas-solid reactions has been designed in order to obtain kinetic data for scale-up purpose. In this bench scale reactor setup, gas and solid reactants can be contacted in a cocurrent and countercurrent manner at high temperatures. Gas and solid sampling can be performed through the reactor bed with their composition profiles determined at steady state. The reactor setup can be used to evaluate and corroborate model parameters accounting for intrinsic reaction rates in both simple and complex gas-solid reaction systems. The moving bed design allows experimentation over a variety of gas and solid compositions in a single experiment unlike differential bed reactors where the gas composition is usually fixed. The data obtained from the reactor can also be used for direct scale-up of designs for moving bed reactors.

Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

NASA Technical Reports Server (NTRS)

Tong, H.; Snow, G. C.; Chu, E. K.; Chang, R. L. S.; Angwin, M. J.; Pessagno, S. L.

1981-01-01

Durable catalytic reactors for advanced gas turbine engines were developed. Objectives were: to evaluate furnace aging as a cost effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1,000 hours of combustion durability, and define a catalytic reactor system with a high probability of successful integration into an automotive gas turbine engine. Fourteen different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel for combustion conditions. Eight additional catalytic reactors were evaluated and one of these was successfully combustion tested on propane fuel. This durability reactor used graded cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

Spinning fluids reactor

DOEpatents

Miller, Jan D; Hupka, Jan; Aranowski, Robert

2012-11-20

A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

Plum Brook Reactor Facility Control Room during Facility Startup

NASA Image and Video Library

1961-02-21

Operators test the National Aeronautics and Space Administration’s (NASA) Plum Brook Reactor Facility systems in the months leading up to its actual operation. The “Reactor On” signs are illuminated but the reactor core was not yet ready for chain reactions. Just a couple weeks after this photograph, Plum Brook Station held a media open house to unveil the 60-megawatt test reactor near Sandusky, Ohio. More than 60 members of the print media and radio and television news services met at the site to talk with community leaders and representatives from NASA and Atomic Energy Commission. The Plum Brook reactor went critical for the first time on the evening of June 14, 1961. It was not until April 1963 that the reactor reached its full potential of 60 megawatts. The reactor control room, located on the second floor of the facility, was run by licensed operators. The operators manually operated the shim rods which adjusted the chain reaction in the reactor core. The regulating rods could partially or completely shut down the reactor. The control room also housed remote area monitoring panels and other monitoring equipment that allowed operators to monitor radiation sensors located throughout the facility and to scram the reactor instantly if necessary. The color of the indicator lights corresponded with the elevation of the detectors in the various buildings. The reactor could also shut itself down automatically if the monitors detected any sudden irregularities.

Research Program of a Super Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

2006-07-01

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is notmore » breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)« less

Experiment for search for sterile neutrino at SM-3 reactor

NASA Astrophysics Data System (ADS)

Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

2016-11-01

In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

Zirconium Hydride Space Power Reactor design.

NASA Technical Reports Server (NTRS)

Asquith, J. G.; Mason, D. G.; Stamp, S.

1972-01-01

The Zirconium Hydride Space Power Reactor being designed and fabricated at Atomics International is intended for a wide range of potential applications. Throughout the program a series of reactor designs have been evaluated to establish the unique requirements imposed by coupling with various power conversion systems and for specific applications. Current design and development emphasis is upon a 100 kilowatt thermal reactor for application in a 5 kwe thermoelectric space power generating system, which is scheduled to be fabricated and ground tested in the mid 70s. The reactor design considerations reviewed in this paper will be discussed in the context of this 100 kwt reactor and a 300 kwt reactor previously designed for larger power demand applications.

A document review to characterize Atomic International SNAP fuels shipped to INEL 1966--1973

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wahnschaffe, S.D.; Lords, R.E.; Kneff, D.W.

1995-09-01

This report provides the results of a document search and review study to obtain information on the spent fuels for the following six Nuclear Auxiliary Power (SNAP) reactor cores now stored at the Idaho National Engineering Laboratory (INEL): SNAP-2 Experimental Reactor, SNAP-2 Development Reactor, SNAP-10A Ground Test Reactor, SNAP-8 Experimental Reactor, SNAP-8 Development Reactor, and Shield Test Reactor. The report also covers documentation on SNAP fuel materials from four in-pile materials tests: NAA-82-1, NAA-115-2, NAA-117-1, and NAA-121. Pieces of these fuel materials are also stored at INEL as part of the SNAP fuel shipments.

A Roadmap of Innovative Nuclear Energy System

NASA Astrophysics Data System (ADS)

Sekimoto, Hiroshi

2017-01-01

Nuclear is a dense energy without CO2 emission. It can be used for more than 100,000 years using fast breeder reactors with uranium from the sea. However, it raises difficult problems associated with severe accidents, spent fuel waste and nuclear threats, which should be solved with acceptable costs. Some innovative reactors have attracted interest, and many designs have been proposed for small reactors. These reactors are considered much safer than conventional large reactors and have fewer technical obstructions. Breed-and-burn reactors have high potential to solve all inherent problems for peaceful use of nuclear energy. However, they have some technical problems with materials. A roadmap for innovative reactors is presented herein.

SNAP 10A FS-3 reactor performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hawley, J.P.; Johnson, R.A.

1966-08-15

SNAP 10FS-3 was the first flight-qualified SNAP reactor system to be operated in a simulated space environment. Prestart-up qualification testing, automatic start-up, endurance period performance, extended operation test and reactor shutdown are described as they affected, or were affected by, overall reactor performance. Performance of the reactor control system and the diagnostic instrumentation is critically evaluted.

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note—A nuclear reactor basically... nuclear reactor and capable of withstanding the operating pressure of the primary coolant. (2) On-line (e...

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note—A nuclear reactor basically... nuclear reactor and capable of withstanding the operating pressure of the primary coolant. (2) On-line (e...

Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

DOEpatents

Hunsbedt, Anstein; Busboom, Herbert J.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

10 CFR 2.621 - Acceptance and docketing of application for early review of site suitability issues in a combined...

Code of Federal Regulations, 2012 CFR

2012-01-01

... Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as...) The Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation... of Nuclear Reactor Regulation, as appropriate, that they are complete. (c) If part one of the...

High yields of hydrogen production from methanol steam reforming with a cross-U type reactor

PubMed Central

Zhang, Shubin; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

2017-01-01

This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance. PMID:29121067

High yields of hydrogen production from methanol steam reforming with a cross-U type reactor.

PubMed

Zhang, Shubin; Zhang, Yufeng; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

2017-01-01

This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance.

Pressurized fluidized bed reactor and a method of operating the same

DOEpatents

Isaksson, J.

1996-02-20

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

Pressurized fluidized bed reactor and a method of operating the same

DOEpatents

Isaksson, Juhani

1996-01-01

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

NASA Astrophysics Data System (ADS)

Kim, Sung-Kyu; Kim, Kwangmin; Park, Minwon; Yu, In-Keun; Lee, Sangjin

2015-11-01

High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

Potential of Electric Power Production from Microbial Fuel Cell (MFC) in Evapotranspiration Reactor for Leachate Treatment Using Alocasia macrorrhiza Plant and Eleusine indica Grass

NASA Astrophysics Data System (ADS)

Zaman, Badrus; Wardhana, Irawan Wisnu

2018-02-01

Microbial fuel cell is one of attractive electric power generator from nature bacterial activity. While, Evapotranspiration is one of the waste water treatment system which developed to eliminate biological weakness that utilize the natural evaporation process and bacterial activity on plant roots and plant media. This study aims to determine the potential of electrical energy from leachate treatment using evapotranspiration reactor. The study was conducted using local plant, namely Alocasia macrorrhiza and local grass, namely Eleusine Indica. The system was using horizontal MFC by placing the cathodes and anodes at different chamber (i.e. in the leachate reactor and reactor with plant media). Carbon plates was used for chatode-anodes material with size of 40 cm x 10 cm x1 cm. Electrical power production was measure by a digital multimeter for 30 days reactor operation. The result shows electric power production was fluctuated during reactor operation from all reactors. The electric power generated from each reactor was fluctuated, but from the reactor using Alocasia macrorrhiza plant reach to 70 μwatt average. From the reactor using Eleusine Indica grass was reached 60 μwatt average. Electric power production fluctuation is related to the bacterial growth pattern in the soil media and on the plant roots which undergo the adaptation process until the middle of the operational period and then in stable growth condition until the end of the reactor operation. The results indicate that the evapotranspiration reactor using Alocasia macrorrhiza plant was 60-95% higher electric power potential than using Eleusine Indica grass in short-term (30-day) operation. Although, MFC system in evapotranspiration reactor system was one of potential system for renewable electric power generation.

Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gougar, Hans David

2015-10-01

The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each ofmore » the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.« less

Preliminary study on aerobic granular biomass formation with aerobic continuous flow reactor

NASA Astrophysics Data System (ADS)

Yulianto, Andik; Soewondo, Prayatni; Handajani, Marissa; Ariesyady, Herto Dwi

2017-03-01

A paradigm shift in waste processing is done to obtain additional benefits from treated wastewater. By using the appropriate processing, wastewater can be turned into a resource. The use of aerobic granular biomass (AGB) can be used for such purposes, particularly for the processing of nutrients in wastewater. During this time, the use of AGB for processing nutrients more reactors based on a Sequencing Batch Reactor (SBR). Studies on the use of SBR Reactor for AGB demonstrate satisfactory performance in both formation and use. SBR reactor with AGB also has been applied on a full scale. However, the use use of SBR reactor still posses some problems, such as the need for additional buffer tank and the change of operation mode from conventional activated sludge to SBR. This gives room for further reactor research with the use of a different type, one of which is a continuous reactor. The purpose of this study is to compare AGB formation using continuous reactor and SBR with same operation parameter. Operation parameter are Organic Loading Rate (OLR) set to 2,5 Kg COD/m3.day with acetate as substrate, aeration rate 3 L/min, and microorganism from Hospital WWTP as microbial source. SBR use two column reactor with volumes 2 m3, and continuous reactor uses continuous airlift reactor, with two compartments and working volume of 5 L. Results from preliminary research shows that although the optimum results are not yet obtained, AGB can be formed on the continuous reactor. When compared with AGB generated by SBR, then the characteristics of granular diameter showed similarities, while the sedimentation rate and Sludge Volume Index (SVI) characteristics showed lower yields.

Assessment of Sensor Technologies for Advanced Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Korsah, Kofi; Kisner, R. A.; Britton Jr., C. L.

This paper provides an assessment of sensor technologies and a determination of measurement needs for advanced reactors (AdvRx). It is a summary of a study performed to provide the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program. The study covered two broad reactor technology categories: High Temperature Reactors and Fast Reactors. The scope of “High temperature reactors” included Gen IV reactors whose coolant exit temperatures exceed ≈650 °C and are moderated (as opposed to fast reactors). To bound the scope formore » fast reactors, this report reviewed relevant operating experience from US-operated Sodium Fast Reactor (SFR) and relevant test experience from the Fast Flux Test Facility (FFTF). For high temperature reactors the study showed that in many cases instrumentation have performed reasonably well in research and demonstration reactors. However, even in cases where the technology is “mature” (such as thermocouples), HTGRs can benefit from improved technologies. Current HTGR instrumentation is generally based on decades-old technology and adapting newer technologies could provide significant advantages. For sodium fast reactors, the study found that several key research needs arise around (1) radiation-tolerant sensor design for in-vessel or in-core applications, where possible non-invasive sensing approaches for key parameters that minimize the need to deploy sensors in-vessel, (2) approaches to exfiltrating data from in-vessel sensors while minimizing penetrations, (3) calibration of sensors in-situ, and (4) optimizing sensor placements to maximize the information content while minimizing the number of sensors needed.« less

Shutdown system for a nuclear reactor

DOEpatents

Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

1984-06-05

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

Reactor vessel support system. [LMFBR

DOEpatents

Golden, M.P.; Holley, J.C.

1980-05-09

A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

NUCLEAR REACTOR AS THE OBJECT OF CONTROL. AUTOMATIC CONTROL OF AIRCRAFT ENGINES . B.S. Voronkev Collection of Articles

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

BS> The dynamics of a power reactor is treated in some detail. Although the reactor is described by a nonlinear differential equation of the seventh order, a two-group approximstion with prompt neutrons and one averaged group of delayed neutrons may be used. When the reactor is in equilibrium, the reactor equation may be linearized in two ways. The effects of positive and negative coefficients of tins of the reactor are discussed. The nonlinear character of the control rods is trested. (D.L.C.)

Shutdown system for a nuclear reactor

DOEpatents

Groh, Edward F.; Olson, Arne P.; Wade, David C.; Robinson, Bryan W.

1984-01-01

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion.

Passive cooling safety system for liquid metal cooled nuclear reactors

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Special features of the inverse-beta-decay reaction proceeding on a proton in a reactor-antineutrino flux

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kopeikin, V. I., E-mail: kopeikin46@yandex.ru; Skorokhvatov, M. D., E-mail: skorokhvatov-md@nrcki.ru

2017-03-15

The evolution of the reactor-antineutrino spectrum and the evolution of the spectrum of positrons from the inverse-beta-decay reaction in the course of reactor operation and after reactor shutdown are considered. The present-day status in determining the initial reactor-antineutrino spectrum on the basis of spectra of beta particles from mixtures of products originating from uranium and plutonium fission is described. A local rise of the experimental spectrum of reactor antineutrinos with respect to the expected spectrum is studied.

An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

NASA Technical Reports Server (NTRS)

Koontz, Steven L.; Davis, Dennis D.; Hansen, Merrill

1988-01-01

A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pan, Paul Y

2010-12-10

An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

Nuclear engine flow reactivity shim control

DOEpatents

Walsh, J.M.

1973-12-11

A nuclear engine control system is provided which automatically compensates for reactor reactivity uncertainties at the start of life and reactivity losses due to core corrosion during the reactor life in gas-cooled reactors. The coolant gas flow is varied automatically by means of specially provided control apparatus so that the reactor control drums maintain a predetermined steady state position throughout the reactor life. This permits the reactor to be designed for a constant drum position and results in a desirable, relatively flat temperature profile across the core. (Official Gazette)

Nuclear Thermal Propulsion: A Joint NASA/DOE/DOD Workshop

NASA Technical Reports Server (NTRS)

Clark, John S. (Editor)

1991-01-01

Papers presented at the joint NASA/DOE/DOD workshop on nuclear thermal propulsion are compiled. The following subject areas are covered: nuclear thermal propulsion programs; Rover/NERVA and NERVA systems; Low Pressure Nuclear Thermal Rocket (LPNTR); particle bed reactor nuclear rocket; hybrid propulsion systems; wire core reactor; pellet bed reactor; foil reactor; Droplet Core Nuclear Rocket (DCNR); open cycle gas core nuclear rockets; vapor core propulsion reactors; nuclear light bulb; Nuclear rocket using Indigenous Martian Fuel (NIMF); mission analysis; propulsion and reactor technology; development plans; and safety issues.

Control rod drive for reactor shutdown

DOEpatents

McKeehan, Ernest R.; Shawver, Bruce M.; Schiro, Donald J.; Taft, William E.

1976-01-20

A means for rapidly shutting down or scramming a nuclear reactor, such as a liquid metal-cooled fast breeder reactor, and serves as a backup to the primary shutdown system. The control rod drive consists basically of an in-core assembly, a drive shaft and seal assembly, and a control drive mechanism. The control rod is driven into the core region of the reactor by gravity and hydraulic pressure forces supplied by the reactor coolant, thus assuring that common mode failures will not interfere with or prohibit scramming the reactor when necessary.

Indirect passive cooling system for liquid metal cooled nuclear reactors

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1990-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Self isolating high frequency saturable reactor

DOEpatents

Moore, James A.

1998-06-23

The present invention discloses a saturable reactor and a method for decoupling the interwinding capacitance from the frequency limitations of the reactor so that the equivalent electrical circuit of the saturable reactor comprises a variable inductor. The saturable reactor comprises a plurality of physically symmetrical magnetic cores with closed loop magnetic paths and a novel method of wiring a control winding and a RF winding. The present invention additionally discloses a matching network and method for matching the impedances of a RF generator to a load. The matching network comprises a matching transformer and a saturable reactor.

Thorium fueled reactor

NASA Astrophysics Data System (ADS)

Sipaun, S.

2017-01-01

Current development in thorium fueled reactors shows that they can be designed to operate in the fast or thermal spectrum. The thorium/uranium fuel cycle converts fertile thorium-232 into fissile uranium-233, which fissions and releases energy. This paper analyses the characteristics of thorium fueled reactors and discusses the thermal reactor option. It is found that thorium fuel can be utilized in molten salt reactors through many configurations and designs. A balanced assessment on the feasibility of adopting one reactor technology versus another could lead to optimized benefits of having thorium resource.

A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

DOE Office of Scientific and Technical Information (OSTI.GOV)

John Darrell Bess

2008-06-01

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, suchmore » as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces.« less

Small space reactor power systems for unmanned solar system exploration missions

NASA Technical Reports Server (NTRS)

Bloomfield, Harvey S.

1987-01-01

A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

A brief history of design studies on innovative nuclear reactors

NASA Astrophysics Data System (ADS)

Sekimoto, Hiroshi

2014-09-01

In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970's the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980's the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

WORKER STACKS GRAPHITE BLOCKS AGAINST INNER SOUTH WALL OF REACTOR. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

WORKER STACKS GRAPHITE BLOCKS AGAINST INNER SOUTH WALL OF REACTOR. INL NEGATIVE NO. 3925. Unknown Photographer, 12/14/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Looking Northeast in Oxide Building at Reactors on Second Floor ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

Looking Northeast in Oxide Building at Reactors on Second Floor Including Reactor One (Left) and Reactor Two (Right) - Hematite Fuel Fabrication Facility, Oxide Building & Oxide Loading Dock, 3300 State Road P, Festus, Jefferson County, MO

Preliminary Comparison of Radioactive Waste Disposal Cost for Fusion and Fission Reactors

NASA Astrophysics Data System (ADS)

Seki, Yasushi; Aoki, Isao; Yamano, Naoki; Tabara, Takashi

1997-09-01

The environmental and economic impact of radioactive waste (radwaste) generated from fusion power reactors using five types of structural materials and a fission reactor has been evaluated and compared. Possible radwaste disposal scenario of fusion radwaste in Japan is considered. The exposure doses were evaluated for the skyshine of gamma-ray during the disposal operation, groundwater migration scenario during the institutional control period of 300 years and future site use scenario after the institutional period. The radwaste generated from a typical light water fission reactor was evaluated using the same methodology as for the fusion reactors. It is found that radwaste from the fusion reactors using F82H and SiC/SiC composites without impurities could be disposed by the shallow land disposal presently applied to the low level waste in Japan. The disposal cost of radwaste from five fusion power reactors and a typical light water reactor were roughly evaluated and compared.

High throughput semiconductor deposition system

DOEpatents

Young, David L.; Ptak, Aaron Joseph; Kuech, Thomas F.; Schulte, Kevin; Simon, John D.

2017-11-21

A reactor for growing or depositing semiconductor films or devices. The reactor may be designed for inline production of III-V materials grown by hydride vapor phase epitaxy (HVPE). The operating principles of the HVPE reactor can be used to provide a completely or partially inline reactor for many different materials. An exemplary design of the reactor is shown in the attached drawings. In some instances, all or many of the pieces of the reactor formed of quartz, such as welded quartz tubing, while other reactors are made from metal with appropriate corrosion resistant coatings such as quartz or other materials, e.g., corrosion resistant material, or stainless steel tubing or pipes may be used with a corrosion resistant material useful with HVPE-type reactants and gases. Using HVPE in the reactor allows use of lower-cost precursors at higher deposition rates such as in the range of 1 to 5 .mu.m/minute.

Nuclear reactor vessel fuel thermal insulating barrier

DOEpatents

Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

2013-03-19

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

Nuclear reactors built, being built, or planned, 1991

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simpson, B.

1992-07-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor ismore » an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

Nuclear component horizontal seismic restraint

DOEpatents

Snyder, Glenn J.

1988-01-01

A nuclear component horizontal seismic restraint. Small gaps limit horizontal displacement of components during a seismic occurrence and therefore reduce dynamic loadings on the free lower end. The reactor vessel and reactor guard vessel use thicker section roll-forged rings welded between the vessel straight shell sections and the bottom hemispherical head sections. The inside of the reactor guard vessel ring forging contains local vertical dovetail slots and upper ledge pockets to mount and retain field fitted and installed blocks. As an option, the horizontal displacement of the reactor vessel core support cone can be limited by including shop fitted/installed local blocks in opposing alignment with the reactor vessel forged ring. Beams embedded in the wall of the reactor building protrude into apertures in the thermal insulation shell adjacent the reactor guard vessel ring and have motion limit blocks attached thereto to provide to a predetermined clearance between the blocks and reactor guard vessel ring.

Safety control circuit for a neutronic reactor

DOEpatents

Ellsworth, Howard C.

2004-04-27

A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, a safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper.

Eastern Europe Research Reactor Initiative nuclear education and training courses - Current activities and future challenges

DOE Office of Scientific and Technical Information (OSTI.GOV)

Snoj, L.; Sklenka, L.; Rataj, J.

2012-07-01

The Eastern Europe Research Reactor Initiative was established in January 2008 to enhance cooperation between the Research Reactors in Eastern Europe. It covers three areas of research reactor utilisation: irradiation of materials and fuel, radioisotope production, neutron beam experiments, education and training. In the field of education and training an EERRI training course was developed. The training programme has been elaborated with the purpose to assist IAEA Member States, which consider building a research reactor (RR) as a first step to develop nuclear competence and infrastructure in the Country. The major strength of the reactor is utilisation of three differentmore » research reactors and a lot of practical exercises. Due to high level of adaptability, the course can be tailored to specific needs of institutions with limited or no access to research reactors. (authors)« less

Eddy Current Flow Measurements in the FFTF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nielsen, Deborah L.; Polzin, David L.; Omberg, Ronald P.

2017-02-02

The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following shut down of the Clinch River Breeder Reactor Plant (CRBRP) project in 1983, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. The FFTF Program provides valuablemore » information for potential follow-on reactor projects in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor control and operations. This report provides HEDL-TC-1344, “ECFM Flow Measurements in the FFTF Using Phase-Sensitive Detectors”, March 1979.« less

Transmutation of actinides in power reactors.

PubMed

Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

2005-01-01

Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

Double-clad nuclear fuel safety rod

DOEpatents

McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

1984-01-01

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

Double-clad nuclear-fuel safety rod

DOEpatents

McCarthy, W.H.; Atcheson, D.B.

1981-12-30

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

NEUTRONIC REACTOR POWER PLANT

DOEpatents

Metcalf, H.E.

1962-12-25

This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2008-11-03

separation list: ! 8 indigenous Indian power reactors ! Fast Breeder test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction...facilities like reprocessing and enrichment plants and breeder reactors could be viewed as providing a significant nonproliferation benefit because the... breeder reactors would support the 2002 U.S. National Strategy to Combat Weapons of Mass Destruction, in which the United States pledged to “continue to

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2008-10-02

8 indigenous Indian power reactors ! Fast Breeder test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction ! Enrichment... breeder reactors could be viewed as providing a significant nonproliferation benefit because the materials produced by these plants are a few steps closer...to potential use in a bomb. In addition, safeguards on enrichment, reprocessing plants, and breeder reactors would support the 2002 U.S. National

BUILDING FOR THE EXPERIMENTAL SWIMMING POOL REACTOR OF 3Mw OF THE JUNTA DE ENERGIA NUCLEAR (in Spanish)

DOE Office of Scientific and Technical Information (OSTI.GOV)

de la Camara, S.N.

1958-10-01

The Spanish experimental swimming pool reactor is constructed on the grounds of the Ciudad Universitaria de Madrid. A general layout of the reactor building and its annexes is given, and the reactor building itself is described. The construction of the reactor building and the characteristics of the annex building are discussed. (J.S.R.)

PRELIMINARY HAZARDS SUMMARY REPORT FOR THE VALLECITOS SUPERHEAT REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murray, J.L.

1961-02-01

BS>The Vallecitos Superheat Reactor (VSR) is a light-watermoderated, thermal-spectrum reactor, cooled by a combination of moderator boiling and forced convection cooling with saturated steam. The reactor core consists of 32 fuel hurdles containing 5300 lb of UO/sub 2/ enriched in U/sub 235/ to 3.6%. The fuel elements are arranged in individual process tubes that direct the cooling steam flow and separate the steam from the water moderator. The reactor vessel is designed for 1250 psig and operates at 960 to 1000 psig. With the reactor operating at 12.5 Mw(t), the maximum fuel cladding temperature is 1250 deg F and themore » cooling steam is superheated to an average temperature of about 810 deg F at 905 psig. Nu clear operation of the reactor is controlled by 12 control rods, actuated by drives mounted on the bottom of the reactor vessel. The water moderator recirculates inside the reactor vessel and through the core region by natural convection. Inherent safety features of the reactor include the negative core reactivity effects upon heating the UO/sub 2/ fuel (Doppler effect), upon increasing the temperature or void content of the moderator in the operating condition, and upon unflooding the fuel process tubes in the hot condition. Snfety features designed into the reactor and plant systems include a system of sensors and devices to detect petentially unsafe operating conditions and to initiate automatically the appropriate countermeasures, a set of fast and reliable control rods for scramming the reactor if a potentially unsafe condition occurs, a manually-actuated liquid neutron poison system, and an emergency cooling system to provide continued steam flow through the reactor core in the event the reactor becomes isolated from either its normal source of steam supply or discharge. The release of radioactivity to unrestricted areas is maintained within permissible limits by monitoring the radioactivity of wastes and controlling their release. The reactor and many of its auxiliaries are housed within a high-integrity essentially leak-tight containment vessel. (auth)« less

Fast formation of aerobic granules by combining strong hydraulic selection pressure with overstressed organic loading rate.

PubMed

Liu, Yong-Qiang; Tay, Joo-Hwa

2015-09-01

The combined strong hydraulic selection pressure (HSP) with overstressed organic loading rate (OLR) as a fast granulation strategy was used to enhance aerobic granulation. To investigate the wide applicability of this strategy to different scenarios and its relevant mechanism, different settling times, different inoculums, different exchange ratios, different reactor configurations, and different shear force were used in this study. It was found that clear granules were formed within 24 h and steady state reached within three days when the fast granulation strategy was used in a lab-scale reactor seeded with well settled activated sludge (Reactor 2). However, granules appeared after 2-week operation and reached steady state after one month at the traditional step-wise decreased settling time from 20 to 2 min with OLR of 6 g COD/L·d (Reactor 1). With the fast granulation strategy, granules appeared within 24 h even with bulking sludge as seed to start up Reactor 3, but 6-day lag phase was observed compared with Reactor 2. Both Reactor 2 and Reactor 3 experienced sigmoidal growth curve in terms of biomass accumulation and granule size increase after granulation. In addition, the reproducible results in pilot-scale reactors (Reactor 5 and Reactor 6) with diameter of 20 cm and height/diameter ratio (H/D) of 4 further proved that reactor configuration and fluid flow pattern had no effect on the aerobic granulation when the fast granulation strategy was employed, but biomass accumulation experienced a short lag phase too in Reactor 5 and Reactor 6. Although overstressed OLR was favorable for fast granulation, it also led to the fluffy granules after around two-week operation. However, the stable 6-month operation of Reactor 3 demonstrated that the rapidly formed granules were able to maintain long-term stability by reducing OLR from 12 g COD/L·d to 6 g COD/L·d. A mechanism of fast granulation with the strategy of combined strong HSP and OLR was proposed to explain results and guide the operation with this fast strategy. Copyright © 2015 Elsevier Ltd. All rights reserved.

Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

NASA Astrophysics Data System (ADS)

Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

2017-01-01

In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

How to Produce a Reactor Neutron Spectrum Using a Proton Accelerator

DOE PAGES

Burns, Kimberly A.; Wootan, David W.; Gates, Robert O.; ...

2015-06-18

A method for reproducing the neutron energy spectrum present in the core of an operating nuclear reactor using an engineered target in an accelerator proton beam is proposed. The protons interact with a target to create neutrons through various (p,n) type reactions. Spectral tailoring of the emitted neutrons can be used to modify the energy of the generated neutron spectrum to represent various reactor spectra. Through the use of moderators and reflectors, the neutron spectrum can be modified to reproduce many different spectra of interest including spectra in small thermal test reactors, large pressurized water reactors, and fast reactors. Themore » particular application of this methodology is the design of an experimental approach for using an accelerator to measure the betas produced during fission to be used to reduce uncertainties in the interpretation of reactor antineutrino measurements. This approach involves using a proton accelerator to produce a neutron field representative of a power reactor, and using this neutron field to irradiate fission foils of the primary isotopes contributing to fission in the reactor, creating unstable, neutron rich fission products that subsequently beta decay and emit electron antineutrinos. A major advantage of an accelerator neutron source over a neutron beam from a thermal reactor is that the fast neutrons can be slowed down or tailored to approximate various power reactor spectra. An accelerator based neutron source that can be tailored to match various reactor neutron spectra provides an advantage for control in studying how changes in the neutron spectra affect parameters such as the resulting fission product beta spectrum.« less

Biological hydrogen production by Clostridium acetobutylicum in an unsaturated flow reactor.

PubMed

Zhang, Husen; Bruns, Mary Ann; Logan, Bruce E

2006-02-01

A mesophilic unsaturated flow (trickle bed) reactor was designed and tested for H2 production via fermentation of glucose. The reactor consisted of a column packed with glass beads and inoculated with a pure culture (Clostridium acetobutylicum ATCC 824). A defined medium containing glucose was fed at a flow rate of 1.6 mL/min (0.096 L/h) into the capped reactor, producing a hydraulic retention time of 2.1 min. Gas-phase H2 concentrations were constant, averaging 74 +/- 3% for all conditions tested. H2 production rates increased from 89 to 220 mL/hL of reactor when influent glucose concentrations were varied from 1.0 to 10.5 g/L. Specific H2 production rate ranged from 680 to 1270 mL/g glucose per liter of reactor (total volume). The H2 yield was 15-27%, based on a theoretical limit by fermentation of 4 moles of H2 from 1 mole of glucose. The major fermentation by-products in the liquid effluent were acetate and butyrate. The reactor rapidly (within 60-72 h) became clogged with biomass, requiring manual cleaning of the system. In order to make long-term operation of the reactor feasible, biofilm accumulation in the reactor will need to be controlled through some process such as backwashing. These tests using an unsaturated flow reactor demonstrate the feasibility of the process to produce high H2 gas concentrations in a trickle-bed type of reactor. A likely application of this reactor technology could be H2 gas recovery from pre-treatment of high carbohydrate-containing wastewaters.

Fail-safe reactivity compensation method for a nuclear reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nygaard, Erik T.; Angelo, Peter L.; Aase, Scott B.

The present invention relates generally to the field of compensation methods for nuclear reactors and, in particular to a method for fail-safe reactivity compensation in solution-type nuclear reactors. In one embodiment, the fail-safe reactivity compensation method of the present invention augments other control methods for a nuclear reactor. In still another embodiment, the fail-safe reactivity compensation method of the present invention permits one to control a nuclear reaction in a nuclear reactor through a method that does not rely on moving components into or out of a reactor core, nor does the method of the present invention rely on themore » constant repositioning of control rods within a nuclear reactor in order to maintain a critical state.« less

HORIZONTAL BOILING REACTOR SYSTEM

DOEpatents

Treshow, M.

1958-11-18

Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

Low-power lead-cooled fast reactor loaded with MOX-fuel

NASA Astrophysics Data System (ADS)

Sitdikov, E. R.; Terekhova, A. M.

2017-01-01

Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

Extension of the TRANSURANUS burnup model to heavy water reactor conditions

NASA Astrophysics Data System (ADS)

Lassmann, K.; Walker, C. T.; van de Laar, J.

1998-06-01

The extension of the light water reactor burnup equations of the TRANSURANUS code to heavy water reactor conditions is described. Existing models for the fission of 235U and the buildup of plutonium in a heavy water reactor are evaluated. In order to overcome the limitations of the frequently used RADAR model at high burnup, a new model is presented. After verification against data for the radial distributions of Xe, Cs, Nd and Pu from electron probe microanalysis, the model is used to analyse the formation of the high burnup structure in a heavy water reactor. The new model allows the analysis of light water reactor fuel rod designs at high burnup in the OECD Halden Heavy Water Reactor.

METHOD FOR SENSING DEGREE OF FLUIDIZATION IN FLUIDIZED BED

DOEpatents

Levey, R.P. Jr.; Fowler, A.H.

1961-12-12

A method is given for detecting, indicating, and controlling the degree of fluidization in a fluid-bed reactor into which powdered material is fed. The method comprises admitting of gas into the reactor, inserting a springsupported rod into the powder bed of the reactor, exciting the rod to vibrate at its resonant frequency, deriving a signal responsive to the amplitude of vibi-ation of the rod and spring, the signal being directiy proportional to the rate of flow of the gas through the reactor, displaying the signal to provide an indication of the degree of fluidization within the reactor, and controlling the rate of gas flow into the reactor until said signal stabilizes at a constant value to provide substantially complete fluidization within the reactor. (AEC)

Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Worrall, Andrew; Todosow, Michael

2016-01-01

Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include:more » increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle performance metrics for a small modular reactor are compared to a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. Metrics performance for a small modular reactor are degraded for mass of spent nuclear fuel and high level waste disposed, mass of depleted uranium disposed, land use per energy generated, and carbon emission per energy generated« less

Aerosol reactor production of uniform submicron powders

NASA Technical Reports Server (NTRS)

Flagan, Richard C. (Inventor); Wu, Jin J. (Inventor)

1991-01-01

A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

Improvement of anaerobic digestion performance by continuous nitrogen removal with a membrane contactor treating a substrate rich in ammonia and sulfide.

PubMed

Lauterböck, B; Nikolausz, M; Lv, Z; Baumgartner, M; Liebhard, G; Fuchs, W

2014-04-01

The effect of reduced ammonia levels on anaerobic digestion was investigated. Two reactors were fed with slaughterhouse waste, one with a hollow fiber membrane contractor for ammonia removal and one without. Different organic loading rates (OLR) and free ammonia and sulfide concentrations were investigated. In the reactor with the membrane contactor, the NH4-N concentration was reduced threefold. At a moderate OLR (3.1 kg chemical oxygen demand - COD/m(3)/d), this reactor performed significantly better than the reference reactor. At high OLR (4.2 kg COD/m(3)/d), the reference reactor almost stopped producing methane (0.01 Nl/gCOD). The membrane reactor also showed a stable process with a methane yield of 0.23 Nl/g COD was achieved. Both reactors had predominantly a hydrogenotrophic microbial consortium, however in the membrane reactor the genus Methanosaeta (acetoclastic) was also detected. In general, all relevant parameters and the methanogenic consortium indicated improved anaerobic digestion of the reactor with the membrane. Copyright © 2014 Elsevier Ltd. All rights reserved.

Control console replacement at the WPI Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

Aerosol reactor production of uniform submicron powders

DOEpatents

Flagan, Richard C.; Wu, Jin J.

1991-02-19

A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

Oxidative coupling of methane using inorganic membrane reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ma, Y.H.; Moser, W.R.; Dixon, A.G.

1995-12-31

The goal of this research is to improve the oxidative coupling of methane in a catalytic inorganic membrane reactor. A specific target is to achieve conversion of methane to C{sub 2} hydrocarbons at very high selectivity and relatively higher yields than in fixed bed reactors by controlling the oxygen supply through the membrane. A membrane reactor has the advantage of precisely controlling the rate of delivery of oxygen to the catalyst. This facility permits balancing the rate of oxidation and reduction of the catalyst. In addition, membrane reactors minimize the concentration of gas phase oxygen thus reducing non selective gasmore » phase reactions, which are believed to be a main route for formation of CO{sub x} products. Such gas phase reactions are a cause for decreased selectivity in oxidative coupling of methane in conventional flow reactors. Membrane reactors could also produce higher product yields by providing better distribution of the reactant gases over the catalyst than the conventional plug flow reactors. Modeling work which aimed at predicting the observed experimental trends in porous membrane reactors was also undertaken in this research program.« less

Stimulation of the hydrolytic stage for biogas production from cattle manure in an electrochemical bioreactor.

PubMed

Samani, Saeed; Abdoli, Mohammad Ali; Karbassi, Abdolreza; Amin, Mohammad Mehdi

Electrical current in the hydrolytic phase of the biogas process might affect biogas yield. In this study, four 1,150 mL single membrane-less chamber electrochemical bioreactors, containing two parallel titanium plates were connected to the electrical source with voltages of 0, -0.5, -1 and -1.5 V, respectively. Reactor 1 with 0 V was considered as a control reactor. The trend of biogas production was precisely checked against pH, oxidation reduction potential and electrical power at a temperature of 37 ± 0.5°C amid cattle manure as substrate for 120 days. Biogas production increased by voltage applied to Reactors 2 and 3 when compared with the control reactor. In addition, the electricity in Reactors 2 and 3 caused more biogas production than Reactor 4. Acetogenic phase occurred more quickly in Reactor 3 than in the other reactors. The obtained results from Reactor 4 were indicative of acidogenic domination and its continuous behavior under electrical stimulation. The results of the present investigation clearly revealed that phasic electrical current could enhance the efficiency of biogas production.

Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

NASA Astrophysics Data System (ADS)

Darmawan, R.

2018-01-01

Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

DOE Office of Scientific and Technical Information (OSTI.GOV)

Widiawati, Nina, E-mail: nina-widiawati28@yahoo.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id

Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uraniummore » fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from −0.6695443 % at BOC to −0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.« less

Five Lectures on Nuclear Reactors Presented at Cal Tech

DOE R&D Accomplishments Database

Weinberg, Alvin M.

1956-02-10

The basic issues involved in the physics and engineering of nuclear reactors are summarized. Topics discussed include theory of reactor design, technical problems in power reactors, physical problems in nuclear power production, and future developments in nuclear power. (C.H.)

10 CFR 2.102 - Administrative review of application.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of...) The Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office... Energy NUCLEAR REGULATORY COMMISSION AGENCY RULES OF PRACTICE AND PROCEDURE Procedure for Issuance...

10 CFR 2.102 - Administrative review of application.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of...) The Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office... Energy NUCLEAR REGULATORY COMMISSION AGENCY RULES OF PRACTICE AND PROCEDURE Procedure for Issuance...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-04-01

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

Navy Nuclear-Powered Surface Ships: Background, Issues, and Options for Congress

DTIC Science & Technology

2010-06-10

scale pressurized water reactors suitable for destroyer-sized vessels or for alternative nuclear power systems using thorium liquid salt technology...or to design a new reactor type potentially using a thorium liquid salt reactor developed for maritime use. The committee recommends an increase of...either using a pressurized water reactor or a thorium liquid salt reactor . (Page 158) Senate The Senate Armed Services Committee, in its report

Reactor monitoring using antineutrino detectors

NASA Astrophysics Data System (ADS)

Bowden, N. S.

2011-08-01

Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactor as part of International Atomic Energy Agency (IAEA) and/or other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway worldwide.

CONVECTION REACTOR

DOEpatents

Hammond, R.P.; King, L.D.P.

1960-03-22

An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

Chip-based device for parallel sorting, amplification, detection, and identification of nucleic acid subsequences

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beer, Neil Reginald; Colston, Jr, Billy W.

An apparatus for chip-based sorting, amplification, detection, and identification of a sample having a planar substrate. The planar substrate is divided into cells. The cells are arranged on the planar substrate in rows and columns. Electrodes are located in the cells. A micro-reactor maker produces micro-reactors containing the sample. The micro-reactor maker is positioned to deliver the micro-reactors to the planar substrate. A microprocessor is connected to the electrodes for manipulating the micro-reactors on the planar substrate. A detector is positioned to interrogate the sample contained in the micro-reactors.

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1974-01-01

The paper summarizes the results obtained in an exploratory evaluation of ceramics for automobile thermal reactors. Candidate ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance, lasting 1100 hours in engine dynamometer tests and for more than 38,600 kilimeters (24,000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1974-01-01

The results obtained in an exploratory evaluation of ceramics for automobile thermal reactors are summarized. Candidate ceramic materials were evaluated in several reactor designs by using both engine-dynamometer and vehicle road tests. Silicon carbide contained in a corrugated-metal support structure exhibited the best performance, lasting 1100 hr in engine-dynamometer tests and more than 38,600 km (24000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as those containing silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

On Study of Application of Micro-reactor in Chemistry and Chemical Field

NASA Astrophysics Data System (ADS)

Zhang, Yunshen

2018-02-01

Serving as a micro-scale chemical reaction system, micro-reactor is characterized by high heat transfer efficiency and mass transfer, strictly controlled reaction time and good safety performance; compared with the traditional mixing reactor, it can effectively shorten reaction time by virtue of these advantages and greatly enhance the chemical reaction conversion rate. However, problems still exist in the process where micro-reactor is used for production in chemistry and chemical field, and relevant researchers are required to optimize and perfect the performance of micro-reactor. This paper analyzes specific application of micro-reactor in chemistry and chemical field.

The results of systems tests of the 500 kV busbar controllable shunting reactor in the Tavricheskaya substation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gusev, S. I.; Karpov, V. N.; Kiselev, A. N.

2009-09-15

The results of systems tests of the 500 kV busbar magnetization-controllable shunting reactor (CSR), set up in the Tavricheskaya substation, including measurements of the quality of the electric power, the harmonic composition of the network currents of the reactor for different values of the reactive power consumed, the determination of the regulating characteristics of the reactor, the speed of response of the shunting reactor in the current and voltage stabilization modes, and also the operation of the reactor under dynamic conditions for different perturbations, are presented. The results obtained are analyzed.

Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

NASA Astrophysics Data System (ADS)

Guidez, Joel; Saturnin, Anne

2017-11-01

During the operation of a nuclear reactor, the external individual doses received by the personnel are measured and recorded, in conformity with the regulations in force. The sum of these measurements enables an evaluation of the annual collective dose expressed in man·Sv/year. This information is a useful tool when comparing the different design types and reactors. This article discusses the evolution of the collective dose for several types of reactors, mainly based on publications from the NEA and the IAEA. The spread of good practices (optimization of working conditions and of the organization, sharing of lessons learned, etc.) and ongoing improvements in reactor design have meant that over time, the doses of various origins received by the personnel have decreased. In the case of sodium-cooled fast reactors (SFRs), the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction). From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

454 pyrosequencing analyses of bacterial and archaeal richness in 21 full-scale biogas digesters.

PubMed

Sundberg, Carina; Al-Soud, Waleed A; Larsson, Madeleine; Alm, Erik; Yekta, Sepehr S; Svensson, Bo H; Sørensen, Søren J; Karlsson, Anna

2013-09-01

The microbial community of 21 full-scale biogas reactors was examined using 454 pyrosequencing of 16S rRNA gene sequences. These reactors included seven (six mesophilic and one thermophilic) digesting sewage sludge (SS) and 14 (ten mesophilic and four thermophilic) codigesting (CD) various combinations of wastes from slaughterhouses, restaurants, households, etc. The pyrosequencing generated more than 160,000 sequences representing 11 phyla, 23 classes, and 95 genera of Bacteria and Archaea. The bacterial community was always both more abundant and more diverse than the archaeal community. At the phylum level, the foremost populations in the SS reactors included Actinobacteria, Proteobacteria, Chloroflexi, Spirochetes, and Euryarchaeota, while Firmicutes was the most prevalent in the CD reactors. The main bacterial class in all reactors was Clostridia. Acetoclastic methanogens were detected in the SS, but not in the CD reactors. Their absence suggests that methane formation from acetate takes place mainly via syntrophic acetate oxidation in the CD reactors. A principal component analysis of the communities at genus level revealed three clusters: SS reactors, mesophilic CD reactors (including one thermophilic CD and one SS), and thermophilic CD reactors. Thus, the microbial composition was mainly governed by the substrate differences and the process temperature. © 2013 Federation of European Microbiological Societies. Published by John Wiley & Sons Ltd. All rights reserved.

Application of a novel type impinging streams reactor in solid-liquid enzyme reactions and modeling of residence time distribution using GDB model.

PubMed

Fatourehchi, Niloufar; Sohrabi, Morteza; Dabir, Bahram; Royaee, Sayed Javid; Haji Malayeri, Adel

2014-02-05

Solid-liquid enzyme reactions constitute important processes in biochemical industries. The isomerization of d-glucose to d-fructose, using the immobilized glucose isomerase (Sweetzyme T), as a typical example of solid-liquid catalyzed reactions has been carried out in one stage and multi-stage novel type of impinging streams reactors. Response surface methodology was applied to determine the effects of certain pertinent parameters of the process namely axial velocity (A), feed concentration (B), nozzles' flow rates (C) and enzyme loading (D) on the performance of the apparatus. The results obtained from the conversion of glucose in this reactor were much higher than those expected in conventional reactors, while residence time was decreased dramatically. Residence time distribution (RTD) in a one-stage impinging streams reactor was investigated using colored solution as the tracer. The results showed that the flow pattern in the reactor was close to that in a continuous stirred tank reactor (CSTR). Based on the analysis of flow region in the reactor, gamma distribution model with bypass (GDB) was applied to study the RTD of the reactor. The results indicated that RTD in the impinging streams reactor could be described by the latter model. Copyright © 2013 Elsevier Inc. All rights reserved.

Enhanced biodegradation of hexachlorocyclohexane in upflow anaerobic sludge blanket reactor using methanol as an electron donor.

PubMed

Bhatt, Praveena; Kumar, M Suresh; Mudliar, Sandeep; Chakrabarti, Tapan

2008-05-01

Anaerobic dechlorination of technical grade hexachlorocyclohexane (THCH) was studied in a continuous upflow anaerobic sludge blanket (UASB) reactor with methanol as a supplementary substrate and electron donor. A reactor without methanol served as the experimental control. The inlet feed concentration of THCH in both the experimental and the control UASB reactor was 100 mg l(-1). After 60 days of continuous operation, the removal of THCH was >99% in the methanol-supplemented reactor as compared to 20-35% in the control reactor. THCH was completely dechlorinated in the methanol fed reactor at 48 h HRT after 2 months of continuous operation. This period was also accompanied by increase in biomass in the reactor, which was not observed in the experimental control. Batch studies using other supplementary substrates as well as electron donors namely acetate, butyrate, formate and ethanol showed lower % dechlorination (<85%) and dechlorination rates (<3 mg g(-1)d(-1)) as compared to methanol (98%, 5 mg g(-1)d(-1)). The optimum concentration of methanol required, for stable dechlorination of THCH (100 mg l(-1)) in the UASB reactor, was found to be 500 mg l(-1). Results indicate that addition of methanol as electron donor enhances dechlorination of THCH at high inlet concentration, and is also required for stable UASB reactor performance.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nygaard, E. T.; Williams, M. M. R.; Angelo, P. L.

Babcock and Wilcox Technical Services Group (B and W) has identified aqueous homogeneous reactors (AHRs) as a technology well suited to produce the medical isotope molybdenum 99 (Mo-99). AHRs have never been specifically designed or built for this specialized purpose. However, AHRs have a proven history of being safe research reactors. In fact, in 1958, AHRs had 'a longer history of operation than any other type of research reactor using enriched fuel' and had 'experimentally demonstrated to be among the safest of all various type of research reactor now in use [1].' A 'Level 1' model representing B and W'smore » proposed Medical Isotope Production System (MIPS) reactor has been developed. The Level 1 model couples a series of differential equations representing neutronics, temperature, and voiding. Neutronics are represented by point reactor kinetics while temperature and voiding terms are axially varying (one-dimensional). While this model was developed specifically for the MIPS reactor, its applicability to the Japanese TRACY reactor was assessed. The results from the Level 1 model were in good agreement with TRACY experimental data and found to be conservative over most of the time domains considered. The Level 1 model was used to study the MIPS reactor. An analysis showed the Level 1 model agreed well with a more complex computational model of the MIPS reactor (a FETCH model). Finally, a significant reactivity insertion was simulated with the Level 1 model to study the MIPS reactor's time-dependent response. (authors)« less

Consumption of the electric power inside silent discharge reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yehia, Ashraf, E-mail: yehia30161@yahoo.com

An experimental study was made in this paper to investigate the relation between the places of the dielectric barriers, which cover the surfaces of the electrodes in the coaxial cylindrical reactors, and the rate of change of the electric power that is consumed in forming silent discharges. Therefore, silent discharges have been formed inside three coaxial cylindrical reactors. The dielectric barriers in these reactors were pasted on both the internal surface of the outer electrode in the first reactor and the external surface of the inner electrode in the second reactor as well as the surfaces of the two electrodesmore » in the third reactor. The reactor under study has been fed by atmospheric air that flowed inside it with a constant rate at normal temperature and pressure, in parallel with the application of a sinusoidal ac voltage between the electrodes of the reactor. The electric power consumed in forming the silent discharges inside the three reactors was measured as a function of the ac peak voltage. The validity of the experimental results was investigated by applying Manley's equation on the same discharge conditions. The results have shown that the rate of consumption of the electric power relative to the ac peak voltage per unit width of the discharge gap improves by a ratio of either 26.8% or 80% or 128% depending on the places of the dielectric barriers that cover the surfaces of the electrodes inside the three reactors.« less

Reactor Operations Monitoring System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hart, M.M.

1989-01-01

The Reactor Operations Monitoring System (ROMS) is a VME based, parallel processor data acquisition and safety action system designed by the Equipment Engineering Section and Reactor Engineering Department of the Savannah River Site. The ROMS will be analyzing over 8 million signal samples per minute. Sixty-eight microprocessors are used in the ROMS in order to achieve a real-time data analysis. The ROMS is composed of multiple computer subsystems. Four redundant computer subsystems monitor 600 temperatures with 2400 thermocouples. Two computer subsystems share the monitoring of 600 reactor coolant flows. Additional computer subsystems are dedicated to monitoring 400 signals from assortedmore » process sensors. Data from these computer subsystems are transferred to two redundant process display computer subsystems which present process information to reactor operators and to reactor control computers. The ROMS is also designed to carry out safety functions based on its analysis of process data. The safety functions include initiating a reactor scram (shutdown), the injection of neutron poison, and the loadshed of selected equipment. A complete development Reactor Operations Monitoring System has been built. It is located in the Program Development Center at the Savannah River Site and is currently being used by the Reactor Engineering Department in software development. The Equipment Engineering Section is designing and fabricating the process interface hardware. Upon proof of hardware and design concept, orders will be placed for the final five systems located in the three reactor areas, the reactor training simulator, and the hardware maintenance center.« less

A brief history of design studies on innovative nuclear reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com

2014-09-30

In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USAmore » and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.« less

A new safety channel based on ¹⁷N detection in research reactors.

PubMed

Seyfi, Somayye; Gharib, Morteza

2015-10-01

Tehran research reactor (TRR) is a representative of pool type research reactors using light water, as coolant and moderator. This reactor is chosen as a prototype to demonstrate and prove the feasibility of (17)N detection as a new redundant channel for reactor power measurement. In TRR, similar to other pool type reactors, neutron detectors are immersed in the pool around the core as the main power measuring devices. In the present article, a different approach, using out of water neutron detector, is employed to measure reactor power. This new method is based on (17)O (n,p) (17)N reaction taking place inside the core and subsequent measurement of delayed neutrons emitted due to (17)N disintegration. Count and measurement of neutrons around outlet water pipe provides a reliable redundant safety channel to measure reactor power. Results compared with other established channels indicate a good agreement and shows a linear interdependency with true thermal power. Safety of reactor operation is improved with installation & use of this new power measuring channel. The new approach may equally serve well as a redundant channel in all other types of reactors having coolant comprised of oxygen in its molecular constituents. Contrary to existing channels, this one is totally out of water and thus is an advantage over current instrumentations. It is proposed to employ the same idea on other reactors (nuclear power plants too) to improve safety criteria. Copyright © 2015 Elsevier Ltd. All rights reserved.

75 FR 21046 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-22

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards In accordance with the... on Reactor Safeguards (ACRS) will hold a meeting on May 6-8, 2010, 11545 Rockville Pike, Rockville....: Boiling Water Reactor (BWR) Owners Group (BWROG) Topical Report NEDC-33347P, ``Containment Overpressure...

Thermionic reactors for space nuclear power

NASA Technical Reports Server (NTRS)

Homeyer, W. G.; Merrill, M. H.; Holland, J. W.; Fisher, C. R.; Allen, D. T.

1985-01-01

Thermionic reactor designs for a variety of space power applications spanning the range from 5 kWe to 3 MWe are described. In all of these reactors, nuclear heat is converted directly to electrical energy in thermionic fuel elements (TFEs). A circulating reactor coolant carries heat from the core of TFEs directly to a heat rejection radiator system. The recent design of a thermionic reactor to meet the SP-100 requirements is emphasized. Design studies of reactors at other power levels show that the same TFE can be used over a broad range in power, and that design modifications can extend the range to many megawatts. The design of the SP-100 TFE is similar to that of TFEs operated successfully in test reactors, but with design improvements to extend the operating lifetime to seven years.

Axi-symmetrical flow reactor for .sup.196 Hg photochemical enrichment

DOEpatents

Grossman, Mark W.

1991-01-01

The present invention is directed to an improved photochemical reactor useful for the isotopic enrichment of a predetermined isotope of mercury, especially, .sup.196 Hg. Specifically, two axi-symmetrical flow reactors were constructed according to the teachings of the present invention. These reactors improve the mixing of the reactants during the photochemical enrichment process, affording higher yields of the desired .sup.196 Hg product. Measurements of the variation of yield (Y) and enrichment factor (E) along the flow axis of these reactors indicates very substantial improvement in process uniformity compared to previously used photochemical reactor systems. In one preferred embodiment of the present invention, the photoreactor system was built such that the reactor chamber was removable from the system without disturbing the location of either the photochemical lamp or the filter employed therewith.

Applications of plasma core reactors to terrestrial energy systems

NASA Technical Reports Server (NTRS)

Latham, T. S.; Biancardi, F. R.; Rodgers, R. J.

1974-01-01

Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

Grey water treatment in upflow anaerobic sludge blanket (UASB) reactor at different temperatures.

PubMed

Elmitwalli, Tarek; Otterpohl, Ralf

2011-01-01

The treatment of grey water in two upflow anaerobic sludge blanket (UASB) reactors, operated at different hydraulic retention times (HRTs) and temperatures, was investigated. The first reactor (UASB-A) was operated at ambient temperature (14-25 degrees C) and HRT of 20, 12 and 8 h, while the second reactor (UASB-30) was operated at controlled temperature of 30 degrees C and HRT of 16, 10 and 6 h. The two reactors were fed with grey water from 'Flintenbreite' settlement in Luebeck, Germany. When the grey water was treated in the UASB reactor at 30 degrees C, total chemical oxygen demand (CODt) removal of 52-64% was achieved at HRT between 6 and 16 h, while at lower temperature lower removal (31-41%) was obtained at HRT between 8 and 20 h. Total nitrogen and phosphorous removal in the UASB reactors were limited (22-36 and 10-24%, respectively) at all operational conditions. The results showed that at increasing temperature or decreasing HRT of the reactors, maximum specific methanogenic activity of the sludge in the reactors improved. As the UASB reactor showed a significantly higher COD removal (31-64%) than the septic tank (11-14%) even at low temperature, it is recommended to use UASB reactor instead of septic tank (the most common system) for grey water pre-treatment. Based on the achieved results and due to high peak flow factor, a HRT between 8 and 12 h can be considered the suitable HRT for the UASB reactor treating grey water at temperature 20-30 degrees C, while a HRT of 12-24 h can be applied at temperature lower than 20 degrees C.

Optimization of post-column reactor radius in capillary high performance liquid chromatography Effect of chromatographic column diameter and particle diameter

PubMed Central

Xu, Hongjuan; Weber, Stephen G.

2006-01-01

A post-column reactor consisting of a simple open tube (Capillary Taylor Reactor) affects the performance of a capillary LC in two ways: stealing pressure from the column and adding band spreading. The former is a problem for very small radius reactors, while the latter shows itself for large reactor diameters. We derived an equation that defines the observed number of theoretical plates (Nobs) taking into account the two effects stated above. Making some assumptions and asserting certain conditions led to a final equation with a limited number of variables, namely chromatographic column radius, reactor radius and chromatographic particle diameter. The assumptions and conditions are that the van Deemter equation applies, the mass transfer limitation is for intraparticle diffusion in spherical particles, the velocity is at the optimum, the analyte’s retention factor, k′, is zero, the post-column reactor is only long enough to allow complete mixing of reagents and analytes and the maximum operating pressure of the pumping system is used. Optimal ranges of the reactor radius (ar) are obtained by comparing the number of observed theoretical plates (and theoretical plates per time) with and without a reactor. Results show that the acceptable reactor radii depend on column diameter, particle diameter, and maximum available pressure. Optimal ranges of ar become narrower as column diameter increases, particle diameter decreases or the maximum pressure is decreased. When the available pressure is 4000 psi, a Capillary Taylor Reactor with 12 μm radius is suitable for all columns smaller than 150 μm (radius) packed with 2–5 μm particles. For 1 μm packing particles, only columns smaller than 42.5 μm (radius) can be used and the reactor radius needs to be 5 μm. PMID:16494886

Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor.

PubMed

Ajijul Hoq, M; Malek Soner, M A; Salam, M A; Haque, M M; Khanom, Salma; Fahad, S M

2017-12-01

The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

Validation of large-scale, monochromatic UV disinfection systems for drinking water using dyed microspheres.

PubMed

Blatchley, E R; Shen, C; Scheible, O K; Robinson, J P; Ragheb, K; Bergstrom, D E; Rokjer, D

2008-02-01

Dyed microspheres have been developed as a new method for validation of ultraviolet (UV) reactor systems. When properly applied, dyed microspheres allow measurement of the UV dose distribution delivered by a photochemical reactor for a given operating condition. Prior to this research, dyed microspheres had only been applied to a bench-scale UV reactor. The goal of this research was to extend the application of dyed microspheres to large-scale reactors. Dyed microsphere tests were conducted on two prototype large-scale UV reactors at the UV Validation and Research Center of New York (UV Center) in Johnstown, NY. All microsphere tests were conducted under conditions that had been used previously in biodosimetry experiments involving two challenge bacteriophage: MS2 and Qbeta. Numerical simulations based on computational fluid dynamics and irradiance field modeling were also performed for the same set of operating conditions used in the microspheres assays. Microsphere tests on the first reactor illustrated difficulties in sample collection and discrimination of microspheres against ambient particles. Changes in sample collection and work-up were implemented in tests conducted on the second reactor that allowed for improvements in microsphere capture and discrimination against the background. Under these conditions, estimates of the UV dose distribution from the microspheres assay were consistent with numerical simulations and the results of biodosimetry, using both challenge organisms. The combined application of dyed microspheres, biodosimetry, and numerical simulation offers the potential to provide a more in-depth description of reactor performance than any of these methods individually, or in combination. This approach also has the potential to substantially reduce uncertainties in reactor validation, thereby leading to better understanding of reactor performance, improvements in reactor design, and decreases in reactor capital and operating costs.

Synthesis of N-graphene using microwave plasma-based methods

NASA Astrophysics Data System (ADS)

Dias, Ana; Tatarova, Elena; Henriques, Julio; Dias, Francisco; Felizardo, Edgar; Abrashev, Miroslav; Bundaleski, Nenad; Cvelbar, Uros

2016-09-01

In this work a microwave atmospheric plasma driven by surface waves is used to produce free-standing graphene sheets (FSG). Carbonaceous precursors are injected into a microwave plasma environment, where decomposition processes take place. The transport of plasma generated gas-phase carbon atoms and molecules into colder zones of plasma reactor results in carbon nuclei formation. The main part of the solid carbon is gradually carried from the ``hot'' plasma zone into the outlet plasma stream where carbon nanostructures assemble and grow. Subsequently, the graphene sheets have been N-doped using a N2-Ar large-scale remote plasma treatment, which consists on placing the FSG on a substrate in a remote zone of the N2-Ar plasma. The samples were treated with different compositions of N2-Ar gas mixtures, while maintaining 1 mbar pressure in the chamber and a power applied of 600 W. The N-doped graphene sheets were characterized by scanning and by high-resolution transmission electron microscopy, X-ray photoelectron spectroscopy and Raman spectroscopy. Plasma characterization was also performed by optical emission spectroscopy. Work partially funded by Portuguese FCT - Fundacao para a Ciencia e a Tecnologia, under grant SFRH/BD/52413/2013 (PD-F APPLAuSE).

BOILING REACTORS

DOEpatents

Untermyer, S.

1962-04-10

A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

Period meter for reactors

DOEpatents

Rusch, Gordon K.

1976-01-06

An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

PBF Reactor Building (PER620). After lowering reactor vessel onto blocks, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). After lowering reactor vessel onto blocks, it is rolled on logs into PBF. Metal framework under vessel is handling device. Various penetrations in reactor bottom were for instrumentation, poison injection, drains. Large one, below center "manhole" was for primary coolant. Photographer: Larry Page. Date: February 13, 1970. INEEL negative no. 70-736 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Liquid metal cooled nuclear reactor plant system

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1993-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

Differences in foot self-care and lifestyle between men and women with diabetes mellitus.

PubMed

Rossaneis, Mariana Angela; Haddad, Maria do Carmo Fernandez Lourenço; Mathias, Thaís Aidar de Freitas; Marcon, Sonia Silva

2016-08-15

to investigate differences with regard to foot self-care and lifestyle between men and women with diabetes mellitus. cross-sectional study conducted in a sample of 1,515 individuals with diabetes mellitus aged 40 years old or older. Poisson regression models were used to identity differences in foot self-care deficit and lifestyle between sexes, adjusting for socioeconomic and clinical characteristics, smoking and alcohol consumption. foot self-care deficit, characterized by not regularly drying between toes; not regularly checking feet; walking barefoot; poor hygiene and inappropriately trimmed nails, was significantly higher among men, though men presented a lower prevalence of feet scaling and use of inappropriate shoes when compared to women. With regard to lifestyle, men presented less healthy habits, such as not adhering to a proper diet and taking laboratory exams to check for lipid profile at the frequency recommended. the nursing team should take into account gender differences concerning foot self-care and lifestyle when implementing educational activities and interventions intended to decrease risk factors for foot ulceration. investigar as diferenças no autocuidado com os pés e no estilo de vida entre mulheres e homens diabéticos. estudo transversal realizado com uma amostra de 1.515 diabéticos com 40 anos ou mais. Foram utilizados modelos de regressão de Poisson para identificar diferenças entre os sexos na prevalência de déficit de autocuidado com os pés e no estilo de vida, ajustando-se por características socioeconômicas, clínicas, tabagismo e alcoolismo. a prevalência de déficit de autocuidado com os pés, caracterizada por baixa frequência de secagem dos espaços interdigitais; da não avaliação periódica dos pés; do hábito de andar descalço; de higiene insatisfatória e corte inadequado de unhas foi significativamente maior entre os homens. Contudo, eles apresentaram menor prevalência na prática de escaldar os pés e no uso de calçados inadequados em comparação às mulheres. Em relação ao estilo de vida, os homens também apresentaram comportamentos menos saudáveis pois tem significativamente menor controle alimentar e não realizam os exames laboratoriais referentes ao perfil lipídico na frequência recomendada. considerar as diferenças de gênero no autocuidado com os pés e no estilo de vida permite à equipe de enfermagem direcionar atividades educacionais e intervenções nos fatores de risco à ulceração dos pés. investigar las diferencias en el autocuidado de los pies y estilo de vida entre mujeres y hombres diabéticos. estudio transversal realizado con una muestra de 1.515 diabéticos con 40 años o más. Fueron utilizados los modelos de regresión de Poisson para identificar diferencias entre los sexos en la prevalencia de déficit de autocuidado de los pies y estilo de vida, ajustándolas por características socioeconómicas, clínicas, tabaquismo y alcoholismo. la prevalencia de déficit de autocuidado de los pies, caracterizada por: baja frecuencia de secado de los espacios interdigitales; falta de evaluación periódica de los pies; hábito de andar descalzo; higiene insatisfactoria; y, corte inadecuado de uñas, fue significativamente mayor entre los hombre. Sin embargo, estos presentaron menor prevalencia en la práctica de escaldar los pies y en el uso de calzados inadecuados en comparación a las mujeres. En relación al estilo de vida, los hombres también presentaron comportamientos menos saludables ya que tienen significativamente menor control alimentario y no realizan los exámenes de laboratorio referentes al perfil lipídico, con la frecuencia recomendada. considerar las diferencias de género en el autocuidado de los pies y en el estilo de vida, le permite al equipo de enfermería realizar actividades educacionales e intervenciones en los factores de riesgo para la ulceración de los pies.

Modelling of the anti-neutrino production and spectra from a Magnox reactor

NASA Astrophysics Data System (ADS)

Mills, Robert W.; Mountford, David J.; Coleman, Jonathon P.; Metelko, Carl; Murdoch, Matthew; Schnellbach, Yan-Jie

2018-01-01

The anti-neutrino source properties of a fission reactor are governed by the production and beta decay of the radionuclides present and the summation of their individual anti-neutrino spectra. The fission product radionuclide production changes during reactor operation and different fissioning species give rise to different product distributions. It is thus possible to determine some details of reactor operation, such as power, from the anti-neutrino emission to confirm safeguards records. Also according to some published calculations, it may be feasible to observe different anti-neutrino spectra depending on the fissile contents of the reactor fuel and thus determine the reactor's fissile material inventory during operation which could considerable improve safeguards. In mid-2014 the University of Liverpool deployed a prototype anti-neutrino detector at the Wylfa R1 station in Anglesey, United Kingdom based upon plastic scintillator technology developed for the T2K project. The deployment was used to develop the detector electronics and software until the reactor was finally shutdown in December 2015. To support the development of this detector technology for reactor monitoring and to understand its capabilities, the National Nuclear Laboratory modelled this graphite moderated and natural uranium fuelled reactor with existing codes used to support Magnox reactor operations and waste management. The 3D multi-physics code PANTHER was used to determine the individual powers of each fuel element (8×6152) during the year and a half period of monitoring based upon reactor records. The WIMS/TRAIL/FISPIN code route was then used to determine the radionuclide inventory of each nuclide on a daily basis in each element. These nuclide inventories were then used with the BTSPEC code to determine the anti-neutrino spectra and source strength using JEFF-3.1.1 data. Finally the anti-neutrino source from the reactor for each day during the year and a half of monitored reactor operation was calculated. The results of the preliminary calculations are shown and limitations in the methods and data discussed.

NASA Reactor Facility Hazards Summary. Volume 1

NASA Technical Reports Server (NTRS)

1959-01-01

The Lewis Research Center of the National Aeronautics and Space Administration proposes to build a nuclear research reactor which will be located in the Plum Brook Ordnance Works near Sandusky, Ohio. The purpose of this report is to inform the Advisory Committee on Reactor Safeguards of the U. S. Atomic Energy Commission in regard to the design Lq of the reactor facility, the characteristics of the site, and the hazards of operation at this location. The purpose of this research reactor is to make pumped loop studies of aircraft reactor fuel elements and other reactor components, radiation effects studies on aircraft reactor materials and equipment, shielding studies, and nuclear and solid state physics experiments. The reactor is light water cooled and moderated of the MTR-type with a primary beryllium reflector and a secondary water reflector. The core initially will be a 3 by 9 array of MTR-type fuel elements and is designed for operation up to a power of 60 megawatts. The reactor facility is described in general terms. This is followed by a discussion of the nuclear characteristics and performance of the reactor. Then details of the reactor control system are discussed. A summary of the site characteristics is then presented followed by a discussion of the larger type of experiments which may eventually be operated in this facility. The considerations for normal operation are concluded with a proposed method of handling fuel elements and radioactive wastes. The potential hazards involved with failures or malfunctions of this facility are considered in some detail. These are examined first from the standpoint of preventing them or minimizing their effects and second from the standpoint of what effect they might have on the reactor facility staff and the surrounding population. The most essential feature of the design for location at the proposed site is containment of the maximum credible accident.

Investigation of the effects of radiolytic-gas bubbles on the long-term operation of solution reactors for medical-isotope production

NASA Astrophysics Data System (ADS)

Souto Mantecon, Francisco Javier

One of the most common and important medical radioisotopes is 99Mo, which is currently produced using the target irradiation technology in heterogeneous nuclear reactors. The medical isotope 99Mo can also be produced from uranium fission using aqueous homogeneous solution reactors. In solution reactors, 99Mo is generated directly in the fuel solution, resulting in potential advantages when compared with the target irradiation process in heterogeneous reactors, such as lower reactor power, less waste heat, and reduction by a factor of about 100 in the generation of spent fuel. The commercial production of medical isotopes in solution reactors requires steady-state operation at about 200 kW. At this power regime, the formation of radiolytic-gas bubbles creates a void volume in the fuel solution that introduces a negative coefficient of reactivity, resulting in power reduction and instabilities that may impede reactor operation for medical-isotope production. A model has been developed considering that reactivity effects are due to the increase in the fuel-solution temperature and the formation of radiolytic-gas bubbles. The model has been validated against experimental results from the Los Alamos National Laboratory uranyl fluoride Solution High-Energy Burst Assembly (SHEBA), and the SILENE uranyl nitrate solution reactor, commissioned at the Commissariat a l'Energie Atomique, in Valduc, France. The model shows the feasibility of solution reactors for the commercial production of medical isotopes and reveals some of the important parameters to consider in their design, including the fuel-solution type, 235U enrichment, uranium concentration, reactor vessel geometry, and neutron reflectors surrounding the reactor vessel. The work presented herein indicates that steady-state operation at 200 kW can be achieved with a solution reactor consisting of 120 L of uranyl nitrate solution enriched up to 20% with 235U and a uranium concentration of 145 kg/m3 in a graphite-reflected cylindrical geometry.

The RERTR Program status and progress

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travelli, A.

1995-12-01

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1995 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1994. The revelation that Iraq was on the verge of developing a nuclear weapon at the time of the Gulf War, and that it was planning to do so by extracting HEU from the fuel of its research reactors, has given new impetus and urgency to the RERTR commitment of eliminating HEU use in research and test reactors worldwide.more » Development of advanced LEU research reactor fuels is scheduled to begin in October 1995. The Russian RERTR program, which aims to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels, is now in operation. A Statement of Intent was signed by high US and Chinese officials, endorsing cooperative activities between the RERTR program and Chinese laboratories involved in similar activities. Joint studies of LEU technical feasibility were completed for the SAFARI-I reactor in South Africa and for the ANS reactor in the US. A new study has been initiated for the FRM-II reactor in Germany. Significant progress was made on several aspects of producing {sup 99}Mo from fission targets utilizing LEU instead of HEU. A cooperation agreements is in place with the Indonesian BATAN. The first prototypical irradiation of an LEU metal-foil target for {sup 99}Mo production was accomplished in Indonesia. The TR-2 reactor, in Turkey, began conversion. SAPHIR, in Switzerland, was shut down. LEU fuel fabrication has begun for the conversion of two more US reactors. Twelve foreign reactors and nine domestic reactors have been fully converted. Approximately 60 % of the work required to eliminate the use of HEU in US-supplied research reactors has been accomplished.« less

Thermionic fast spectrum reactor-converter on the basis of multi-cell TFE

NASA Astrophysics Data System (ADS)

Ponomarev-Stepnoi, N. N.; Kompaniets, G. V.; Poliakov, D. N.; Stepennov, B. S.; Andreev, P. V.; Zhabotinsky, E. E.; Nikolaev, Yu. V.; Lapochkin, N. V.

2001-02-01

Today Russian experts have technological experience in development of in-core thermionic converters for reactors of space nuclear power plants. Such a converter contains nuclear fuel inside and really represents a fuel element of a reactor. Two types of reactors can be considered on the basis of these thermionic fuel elements: with thermal or intermediate neutron spectrum, and with fast neutron spectrum. The first type is characterized by the presence of moderator in core that ensures most economical usage of nuclear fuel. The estimation shows that moderated system is the most effective in the power range of about 5 ... 100 kWe. The power systems of higher level are characterized by larger dimensions due to the presence of moderator. The second type of reactor is considered for higher power levels. This power range is about hundreds kWe. Dimensions of the fast reactor and core configuration are determined by the necessity to ensure the required net output power, on the one hand, and the necessity to ensure critical state on the other hand. In the case of using in-core thermionic fuel elements of the specified design, minimal reactor output power is determined by reactor criticality condition, and maximum reactor power output is determined by specifications and launcher capabilities. In the present paper the effective multiplication factor of a fast spectrum reactor on the basis of a multi-cell TFE developed by ``Lutch'' is considered a function of the total number of TFEs in the reactor. The MCU Monte-Carlo code, developed in Russia (Alekseev, et al., 1991), was used for computations. TFE computational models are placed in the nodes of a uniform triangular lattice and surrounded with pressure vessel and a side reflector. Ordinary fuel pins without thermionic converters were used instead of some TFEs to optimize criticality parameters, dimensions and output power of the reactor. General weight parameters of the reactor are presented in the paper. .

Management of fresh water weeds (macrophytes) by vermicomposting using Eisenia fetida.

PubMed

Najar, Ishtiyaq Ahmed; Khan, Anisa B

2013-09-01

In the present study, potential of Eisenia fetida to recycle the different types of fresh water weeds (macrophytes) used as substrate in different reactors (Azolla pinnata reactor, Trapa natans reactor, Ceratophyllum demersum reactor, free-floating macrophytes mixture reactor, and submerged macrophytes mixture reactor) during 2 months experiment is investigated. E. fetida showed significant variation in number and weight among the reactors and during the different fortnights (P <0.05) with maximum in A. pinnata reactor (number 343.3 ± 10.23 %; weight 98.62 ± 4.23 % ) and minimum in submerged macrophytes mixture reactor (number 105 ± 5.77 %; weight 41.07 ± 3.97 % ). ANOVA showed significant variation in cocoon production (F4 = 15.67, P <0.05) and mean body weight (F4 = 13.49, P <0.05) among different reactors whereas growth rate (F3 = 23.62, P <0.05) and relative growth rate (F3 = 4.91, P <0.05) exhibited significant variation during different fortnights. Reactors showed significant variation (P <0.05) in pH, Electrical conductivity (EC), Organic carbon (OC), Organic nitrogen (ON), and C/N ratio during different fortnights with increase in pH, EC, N, and K whereas decrease in OC and C/N ratio. Hierarchical cluster analysis grouped five substrates (weeds) into three clusters-poor vermicompost substrates, moderate vermicompost substrate, and excellent vermicompost substrate. Two principal components (PCs) have been identified by factor analysis with a cumulative variance of 90.43 %. PC1 accounts for 47.17 % of the total variance represents "reproduction factor" and PC2 explaining 43.26 % variance representing "growth factor." Thus, the nature of macrophyte affects the growth and reproduction pattern of E. fetida among the different reactors, further the addition of A. pinnata in other macrophytes reactors can improve their recycling by E. fetida.

76 FR 16842 - Request for a License To Export Reactor Components

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-25

... NUCLEAR REGULATORY COMMISSION Request for a License To Export Reactor Components Pursuant to 10.... Mechanical Corporation. coolant pump 1000 (design) maintenance, and systems, related reactors. operation of AP- equipment, and 1000 (design) spare parts. nuclear reactors. February 10, 2011 February 23, 2011...

76 FR 64126 - Advisory Committee on Reactor Safeguards; Procedures for Meetings

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-17

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Procedures for Meetings.... Nuclear Regulatory Commission's (NRC's) Advisory Committee on Reactor Safeguards (ACRS) pursuant to the... specified in the Federal Register Notice, care of the Advisory Committee on Reactor Safeguards, U.S. Nuclear...

78 FR 35056 - Effectiveness of the Reactor Oversight Process Baseline Inspection Program

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-11

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0125] Effectiveness of the Reactor Oversight Process... the effectiveness of the reactor oversight process (ROP) baseline inspection program with members of... Nuclear Reactor Regulations, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301...

78 FR 67205 - Advisory Committee on Reactor Safeguards; Procedures for Meetings

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-08

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Procedures for Meetings.... Nuclear Regulatory Commission's (NRC's) Advisory Committee on Reactor Safeguards (ACRS) pursuant to the... specified in the Federal Register Notice, care of the Advisory Committee on Reactor Safeguards, U.S. Nuclear...

77 FR 60039 - Non-Power Reactor License Renewal

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-02

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [NRC-2011-0087] RIN 3150-AI96 Non-Power Reactor... the final regulatory basis for rulemaking to streamline non-power reactor license renewal. This final... Reactor (RTR) License Renewal Process. This contemplated rulemaking also recommends conforming changes to...

75 FR 8154 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-23

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards In accordance with the... on Reactor Safeguards (ACRS) will hold a meeting on March 4-6, 2010, 11545 Rockville Pike, Rockville....-12 p.m.: New Advanced Reactor Designs (Open)--The Committee will hear presentations by and hold...

10 CFR 100.3 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... COMMISSION (CONTINUED) REACTOR SITE CRITERIA § 100.3 Definitions. As used in this part: Combined license... power facilities. Exclusion area means that area surrounding the reactor, in which the reactor licensee.... Activities unrelated to operation of the reactor may be permitted in an exclusion area under appropriate...

77 FR 38742 - Non-Power Reactor License Renewal

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-29

...-0087] RIN 3150-AI96 Non-Power Reactor License Renewal AGENCY: Nuclear Regulatory Commission. ACTION... reactors. This contemplated rulemaking would also make conforming changes to address technical issues in existing non-power reactor regulations. The NRC is seeking input from the public, licensees, certificate...

10 CFR 100.3 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... COMMISSION (CONTINUED) REACTOR SITE CRITERIA § 100.3 Definitions. As used in this part: Combined license... power facilities. Exclusion area means that area surrounding the reactor, in which the reactor licensee.... Activities unrelated to operation of the reactor may be permitted in an exclusion area under appropriate...

78 FR 58575 - Review of Experiments for Research Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-24

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0219] Review of Experiments for Research Reactors AGENCY... Commission (NRC) is withdrawing Regulatory Guide (RG) 2.4, ``Review of Experiments for Research Reactors... withdrawing RG 2.4, ``Review of Experiments for Research Reactors,'' (ADAMS Accession No. ML003740131) because...

NEUTRONIC REACTOR SHIELDING

DOEpatents

Borst, L.B.

1961-07-11

A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

5 CFR 5801.102 - Prohibited securities.

Code of Federal Regulations, 2014 CFR

2014-01-01

... licenses for facilities which generate electric energy by means of a nuclear reactor; (2) State or local... reactor or a low-level waste facility; (3) Entities manufacturing or selling nuclear power or test reactors; (4) Architectural-engineering companies providing services relating to a nuclear power reactor...

10 CFR 2.101 - Filing of application.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Nuclear Reactor Regulation, the Director, Office of Nuclear Material Safety and Safeguards, or the... this chapter, see paragraph (g) of this section. (3) If the Director, Office of Nuclear Reactor...) Submit to the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director...

10 CFR 2.101 - Filing of application.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Nuclear Reactor Regulation, the Director, Office of Nuclear Material Safety and Safeguards, or the... this chapter, see paragraph (g) of this section. (3) If the Director, Office of Nuclear Reactor...) Submit to the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director...

5 CFR 5801.102 - Prohibited securities.

Code of Federal Regulations, 2010 CFR

2010-01-01

... licenses for facilities which generate electric energy by means of a nuclear reactor; (2) State or local... reactor or a low-level waste facility; (3) Entities manufacturing or selling nuclear power or test reactors; (4) Architectural-engineering companies providing services relating to a nuclear power reactor...

Reactor performance and microbial community dynamics during anaerobic co-digestion of municipal wastewater sludge with restaurant grease waste at steady state and overloading stages.

PubMed

Razaviarani, Vahid; Buchanan, Ian D

2014-11-01

Linkage between reactor performance and microbial community dynamics was investigated during mesophilic anaerobic co-digestion of restaurant grease waste (GTW) with municipal wastewater sludge (MWS) using 10L completely mixed reactors and a 20day SRT. Test reactors received a mixture of GTW and MWS while control reactors received only MWS. Addition of GTW to the test reactors enhanced the biogas production and methane yield by up to 65% and 120%, respectively. Pyrosequencing revealed that Methanosaeta and Methanomicrobium were the dominant acetoclastic and hydrogenotrophic methanogen genera, respectively, during stable reactor operation. The number of Methanosarcina and Methanomicrobium sequences increased and that of Methanosaeta declined when the proportion of GTW in the feed was increased to cause an overload condition. Under this overload condition, the pH, alkalinity and methane production decreased and VFA concentrations increased dramatically. Candidatus cloacamonas, affiliated within phylum Spirochaetes, were the dominant bacterial genus at all reactor loadings. Copyright © 2014 Elsevier Ltd. All rights reserved.

Modifications to the NRAD Reactor, 1977 to present

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weeks, A.A.; Pruett, D.P.; Heidel, C.C.

1986-01-01

Argonne National Laboratory-West, operated by the University of Chicago, is located near Idaho Falls, ID, on the Idaho National Engineering laboratory Site. ANL-West performs work in support of the Liquid Metal Fast Breeder Reactor Program (LMFBR) sponsored by the United States Department of Energy. The NRAD reactor is located at the Argonne Site within the Hot Fuel Examination Facility/North, a large hot cell facility where both non-destructive and destructive examinations are performed on highly irradiated reactor fuels and materials in support of the LMFBR program. The NRAD facility utilizes a 250-kW TRIGA reactor and is completely dedicated to neutron radiographymore » and the development of radiography techniques. Criticality was first achieved at the NRAD reactor in October of 1977. Since that time, a number of modifications have been implemented to improve operational efficiency and radiography production. This paper describes the modifications and changes that significantly improved operational efficiency and reliability of the reactor and the essential auxiliary reactor systems.« less

Process of simultaneous hydrogen sulfide removal from biogas and nitrogen removal from swine wastewater.

PubMed

Deng, Liangwei; Chen, Huijuan; Chen, Ziai; Liu, Yi; Pu, Xiaodong; Song, Li

2009-12-01

The feasibility of a new flowchart describing simultaneous hydrogen sulfide removal from biogas and nitrogen removal from wastewater was investigated. It took 30 days for the reactor inoculated with aerobic sludge to attain a removal rate of 60% for H(2)S and NO(x)-N simultaneously. It took 34 and 48 days to attain the same removal rate for the reactor without inoculated sludge and the reactor inoculated with anaerobic sludge respectively. The reactor without inoculated sludge still operated successfully, despite requiring a slightly longer startup time. The packing material was capable of enhancing the removal efficiency of reactors. Based on the concentration of NO(x)-N and H(2)S in the effluent, the loading rate and the ability of the system to resist shock loading, the performance of the reactor filled with hollow plastic balls was greater than that of the reactor filled with elastic packing and the reactor filled with Pall rings.

Treatment of oilfield wastewater in moving bed biofilm reactors using a novel suspended ceramic biocarrier.

PubMed

Dong, Zhiyong; Lu, Mang; Huang, Wenhui; Xu, Xiaochun

2011-11-30

In this study, a novel suspended ceramic carrier was prepared, which has high strength, optimum density (close to water), and high porosity. Two different carriers, unmodified and sepiolite-modified suspended ceramic carriers were used to feed two moving bed biofilm reactors (MBBRs) with a filling fraction of 50% to treat oilfield produced water. The hydraulic retention time (HRT) was varied from 36 to 10h. The results, during a monitoring period of 190 days, showed that removal efficiency of chemical oxygen demand was the highest in reactor 3 filled with the sepiolite-modified carriers, followed by reactor 2 filled with the unmodified carriers, with the lowest in reactor 1 (activated sludge reactor), at an HRT of 10h. Similar trends were found in the removal efficiencies of ammonia nitrogen and polycyclic aromatic hydrocarbons. Reactor 3 was more shock resistant than reactors 2 and 1. The results indicate that the suspended ceramic carrier is an excellent MBBR carrier. Copyright © 2011 Elsevier B.V. All rights reserved.

Integrated hydrocarbon reforming system and controls

DOEpatents

Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Thijssen, Johannes; Davis, Robert; Papile, Christopher; Rumsey, Jennifer W.; Longo, Nathan; Cross, III, James C.; Rizzo, Vincent; Kleeburg, Gunther; Rindone, Michael; Block, Stephen G.; Sun, Maria; Morriseau, Brian D.; Hagan, Mark R.; Bowers, Brian

2003-11-04

A hydrocarbon reformer system including a first reactor configured to generate hydrogen-rich reformate by carrying out at least one of a non-catalytic thermal partial oxidation, a catalytic partial oxidation, a steam reforming, and any combinations thereof, a second reactor in fluid communication with the first reactor to receive the hydrogen-rich reformate, and having a catalyst for promoting a water gas shift reaction in the hydrogen-rich reformate, and a heat exchanger having a first mass of two-phase water therein and configured to exchange heat between the two-phase water and the hydrogen-rich reformate in the second reactor, the heat exchanger being in fluid communication with the first reactor so as to supply steam to the first reactor as a reactant is disclosed. The disclosed reformer includes an auxiliary reactor configured to generate heated water/steam and being in fluid communication with the heat exchanger of the second reactor to supply the heated water/steam to the heat exchanger.

Nuclear reactor control column

DOEpatents

Bachovchin, Dennis M.

1982-01-01

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

Standardized reactors for the study of medical biofilms: a review of the principles and latest modifications.

PubMed

Gomes, Inês B; Meireles, Ana; Gonçalves, Ana L; Goeres, Darla M; Sjollema, Jelmer; Simões, Lúcia C; Simões, Manuel

2018-08-01

Biofilms can cause severe problems to human health due to the high tolerance to antimicrobials; consequently, biofilm science and technology constitutes an important research field. Growing a relevant biofilm in the laboratory provides insights into the basic understanding of the biofilm life cycle including responses to antibiotic therapies. Therefore, the selection of an appropriate biofilm reactor is a critical decision, necessary to obtain reproducible and reliable in vitro results. A reactor should be chosen based upon the study goals and a balance between the pros and cons associated with its use and operational conditions that are as similar as possible to the clinical setting. However, standardization in biofilm studies is rare. This review will focus on the four reactors (Calgary biofilm device, Center for Disease Control biofilm reactor, drip flow biofilm reactor, and rotating disk reactor) approved by a standard setting organization (ASTM International) for biofilm experiments and how researchers have modified these standardized reactors and associated protocols to improve the study and understanding of medical biofilms.

Inhibition of nitrification in municipal wastewater-treating photobioreactors: Effect on algal growth and nutrient uptake.

PubMed

Krustok, I; Odlare, M; Truu, J; Nehrenheim, E

2016-02-01

The effect of inhibiting nitrification on algal growth and nutrient uptake was studied in photobioreactors treating municipal wastewater. As previous studies have indicated that algae prefer certain nitrogen species to others, and because nitrifying bacteria are inhibited by microalgae, it is important to shed more light on these interactions. In this study allylthiourea (ATU) was used to inhibit nitrification in wastewater-treating photobioreactors. The nitrification-inhibited reactors were compared to control reactors with no ATU added. Microalgae had higher growth in the inhibited reactors, resulting in a higher chlorophyll a concentration. The species mix also differed, with Chlorella and Scenedesmus being the dominant genera in the control reactors and Cryptomonas and Chlorella dominating in the inhibited reactors. The nitrogen speciation in the reactors after 8 days incubation was also different in the two setups, with N existing mostly as NH4-N in the inhibited reactors and as NO3-N in the control reactors. Copyright © 2015 Elsevier Ltd. All rights reserved.

Effets de l'humidite sur la propagation du delaminage dans un composite carbone/epoxy sollicite en mode mixte I/II

NASA Astrophysics Data System (ADS)

LeBlanc, Luc R.

Les materiaux composites sont de plus en plus utilises dans des domaines tels que l'aerospatiale, les voitures a hautes performances et les equipements sportifs, pour en nommer quelques-uns. Des etudes ont demontre qu'une exposition a l'humidite nuit a la resistance des composites en favorisant l'initiation et la propagation du delaminage. De ces etudes, tres peu traitent de l'effet de l'humidite sur l'initiation du delaminage en mode mixte I/II et aucune ne traite des effets de l'humidite sur le taux de propagation du delaminage en mode mixte I/II dans un composite. La premiere partie de cette these consiste a determiner les effets de l'humidite sur la propagation du delaminage lors d'une sollicitation en mode mixte I/II. Des eprouvettes d'un composite unidirectionnel de carbone/epoxy (G40-800/5276-1) ont ete immergees dans un bain d'eau distillee a 70°C jusqu'a leur saturation. Des essais experimentaux quasi-statiques avec des chargements d'une gamme de mixites des modes I/II (0%, 25%, 50%, 75% et 100%) ont ete executes pour determiner les effets de l'humidite sur la resistance au delaminage du composite. Des essais de fatigue ont ete realises, avec la meme gamme de mixite des modes I/II, pour determiner 1'effet de 1'humidite sur l'initiation et sur le taux de propagation du delaminage. Les resultats des essais en chargement quasi-statique ont demontre que l'humidite reduit la resistance au delaminage d'un composite carbone/epoxy pour toute la gamme des mixites des modes I/II, sauf pour le mode I ou la resistance au delaminage augmente apres une exposition a l'humidite. Pour les chargements en fatigue, l'humidite a pour effet d'accelerer l'initiation du delaminage et d'augmenter le taux de propagation pour toutes les mixites des modes I/II. Les donnees experimentales recueillies ont ete utilisees pour determiner lesquels des criteres de delaminage en statique et des modeles de taux de propagation du delaminage en fatigue en mode mixte I/II proposes dans la litterature representent le mieux le delaminage du composite etudie. Une courbe de regression a ete utilisee pour determiner le meilleur ajustement entre les donnees experimentales et les criteres de delaminage en statique etudies. Une surface de regression a ete utilisee pour determiner le meilleur ajustement entre les donnees experimentales et les modeles de taux de propagation en fatigue etudies. D'apres les ajustements, le meilleur critere de delaminage en statique est le critere B-K et le meilleur modele de propagation en fatigue est le modele de Kenane-Benzeggagh. Afin de predire le delaminage lors de la conception de pieces complexes, des modeles numeriques peuvent etre utilises. La prediction de la longueur de delaminage lors des chargements en fatigue d'une piece est tres importante pour assurer qu'une fissure interlaminaire ne va pas croitre excessivement et causer la rupture de cette piece avant la fin de sa duree de vie de conception. Selon la tendance recente, ces modeles sont souvent bases sur l'approche de zone cohesive avec une formulation par elements finis. Au cours des travaux presentes dans cette these, le modele de progression du delaminage en fatigue de Landry & LaPlante (2012) a ete ameliore en y ajoutant le traitement des chargements en mode mixte I/II et en y modifiant l'algorithme du calcul de la force d'entrainement maximale du delaminage. Une calibration des parametres de zone cohesive a ete faite a partir des essais quasi-statiques experimentaux en mode I et II. Des resultats de simulations numeriques des essais quasi-statiques en mode mixte I/II, avec des eprouvettes seches et humides, ont ete compares avec les essais experimentaux. Des simulations numeriques en fatigue ont aussi ete faites et comparees avec les resultats experimentaux du taux de propagation du delaminage. Les resultats numeriques des essais quasi-statiques et de fatigue ont montre une bonne correlation avec les resultats experimentaux pour toute la gamme des mixites des modes I/II etudiee.

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

DOE PAGES

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.; ...

2016-12-21

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Indah Rosidah, M., E-mail: indah.maymunah@gmail.com; Suud, Zaki, E-mail: szaki@fi.itb.ac.id; Yazid, Putranto Ilham

The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With themore » tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature and high pressure condition for more than ten years.« less

Project Luna Succendo: The Lunar Evolutionary Growth-Optimized (LEGO) Reactor

NASA Astrophysics Data System (ADS)

Bess, John Darrell

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched within lunar shipments from the Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides 5 kWe using a free-piston Stirling space converter. The overall envelope for a single unit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. The subunits can be placed with centerline distances of approximately 0.6 m in a hexagonal-lattice pattern to provide sufficient neutronic coupling while allowing room for heat rejection and interstitial control. A lattice of six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network Future improvements include advances in reactor control methods, fuel form and matrix, determination of shielding requirements, as well as power conversion and heat rejection techniques to generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces such as Mars, other moons, and asteroids.

Apparatus and process for the surface treatment of carbon fibers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paulauskas, Felix Leonard; Ozcan, Soydan; Naskar, Amit K.

A method for surface treating a carbon-containing material in which carbon-containing material is reacted with decomposing ozone in a reactor (e.g., a hollow tube reactor), wherein a concentration of ozone is maintained throughout the reactor by appropriate selection of at least processing temperature, gas stream flow rate, reactor dimensions, ozone concentration entering the reactor, and position of one or more ozone inlets (ports) in the reactor, wherein the method produces a surface-oxidized carbon or carbon-containing material, preferably having a surface atomic oxygen content of at least 15%. The resulting surface-oxidized carbon material and solid composites made therefrom are also described.

Development concept for a small, split-core, heat-pipe-cooled nuclear reactor

NASA Technical Reports Server (NTRS)

Lantz, E.; Breitwieser, R.; Niederauer, G. F.

1974-01-01

There have been two main deterrents to the development of semiportable nuclear reactors. One is the high development costs; the other is the inability to satisfy with assurance the questions of operational safety. This report shows how a split-core, heat-pipe cooled reactor could conceptually eliminate these deterrents, and examines and summarizes recent work on split-core, heat-pipe reactors. A concept for a small reactor that could be developed at a comparatively low cost is presented. The concept would extend the technology of subcritical radioisotope thermoelectric generators using 238 PuO2 to the evolution of critical space power reactors using 239 PuO2.

Summary of the Workshop on Molten Salt Reactor Technologies Commemorating the 50th Anniversary of the Startup of the Molten Salt Reactor Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Betzler, Benjamin R; Mays, Gary T

2016-01-01

A workshop on Molten Salt Reactor (MSR) technologies commemorating the 50th anniversary of the Molten Salt Reactor Experiment (MSRE) was held at Oak Ridge National Laboratory on October 15 16, 2015. The MSRE represented a pioneering experiment that demonstrated an advanced reactor technology: the molten salt eutectic-fueled reactor. A multinational group of more than 130 individuals representing a diverse set of stakeholders gathered to discuss the historical, current, and future technical challenges and paths to deployment of MSR technology. This paper provides a summary of the key messages from this workshop.

Fast-acting nuclear reactor control device

DOEpatents

Kotlyar, Oleg M.; West, Phillip B.

1993-01-01

A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

Exploratory development of a glass ceramic automobile thermal reactor. [anti-pollution devices

NASA Technical Reports Server (NTRS)

Gould, R. E.; Petticrew, R. W.

1973-01-01

This report summarizes the design, fabrication and test results obtained for glass-ceramic (CER-VIT) automotive thermal reactors. Several reactor designs were evaluated using both engine-dynamometer and vehicle road tests. A maximum reactor life of about 330 hours was achieved in engine-dynamometer tests with peak gas temperatures of about 1065 C (1950 F). Reactor failures were mechanically induced. No evidence of chemical degradation was observed. It was concluded that to be useful for longer times, the CER-VIT parts would require a mounting system that was an improvement over those tested in this program. A reactor employing such a system was designed and fabricated.

Exploratory evaluation of ceramics for automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1972-01-01

An exploratory evaluation of ceramics for automobile thermal reactors was conducted. Potential ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance lasting over 800 hours in engine dynamometer tests and over 15,000 miles (24,200 km) of vehicle road tests. Reactors containing glass-ceramic components did not perform as well as silicon carbide. But the glass-ceramics still offer good potential for reactor use. The results of this study are considered to be a reasonable demonstration of the potential use of ceramics in thermal reactors.

Nuclear reactor overflow line

DOEpatents

Severson, Wayne J.

1976-01-01

The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-07-01

Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Bergeron, A.; Dionne, B.

2016-01-01

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

Removal of slowly biodegradable COD in combined thermophilic UASB and MBBR systems.

PubMed

Ji, M; Yu, J; Chen, H; Yue, P L

2001-09-01

Starch, cellulose and polyvinyl alcohol (PVA) are common substrates of the slowly biodegradable COD (SBCOD) in industrial wastewaters. Removal of the individual and mixed SbCOD substrates was investigated in a combined system of thermophilic upflow anaerobic sludge blanket (TUASB) reactor (55 degrees C) and aerobic moving bed biofilm reactor (MBBR). The removal mechanisms of the three SBCOD substrates were quite different. Starch-COD was almost equally utilized and removed in the two reactors. Cellulose-COD was completely (97-98%) removed from water in the TUASB reactor by microbial entrapment and sedimentation of the cellulose fibers. PVA alone was hardly biodegraded and removed by the combined reactors. However, PVA-COD could be removed to some extent in a binary solution of starch (77%) plus PVA (23%). The PVA macromolecules in the binary solution actually affected the microbial activity in the TUASB reactor resulting accumulation of volatile fatty acids, which shifted the overall COD removal from the TUASB to the MBBR reactor where SBCOD including PVA-COD was removed. Since the three SBCOD substrates were removed by different mechanisms, the combined reactors showed a better and more stable performance than individual reactors.

Impact of non-ionic surfactant on the long-term development of lab-scale-activated sludge bacterial communities.

PubMed

Lozada, Mariana; Basile, Laura; Erijman, Leonardo

2007-01-01

The development of bacterial communities in replicate lab-scale-activated sludge reactors degrading a non-ionic surfactant was evaluated by statistical analysis of denaturing gradient gel electrophoresis (DGGE) fingerprints. Four sequential batch reactors were fed with synthetic sewage, two of which received, in addition, 0.01% of nonylphenol ethoxylates (NPE). The dynamic character of bacterial community structure was confirmed by the differences in species composition among replicate reactors. Measurement of similarities between reactors was obtained by pairwise similarity analysis using the Bray Curtis coefficient. The group of NPE-amended reactors exhibited the highest similarity values (Sjk=0.53+/-0.03), indicating that the bacterial community structure of NPE-amended reactors was better replicated than control reactors (Sjk=0.36+/-0.04). Replicate NPE-amended reactors taken at different times of operation clustered together, whereas analogous relations within the control reactor cluster were not observed. The DGGE pattern of isolates grown in conditioned media prepared with media taken at the end of the aeration cycle grouped separately from other conditioned and synthetic media regardless of the carbon source amendment, suggesting that NPE degradation residuals could have a role in the shaping of the community structure.

The WPI reactor-readying for the next generation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bobek, L.M.

1993-01-01

Built in 1959, the 10-kW open-pool nuclear training reactor at Worcester Polytechnic Institute (WPI) was one of the first such facilities in the nation located on a university campus. Since then, the reactor and its related facilities have been used to train two generations of nuclear engineers and scientists for the nuclear industry. With the use of nuclear technology playing an increasing role in many segments of the economy, WPI with its nuclear reactor facility is committed to continuing its mission of training future nuclear engineers and scientists. The WPI reactor includes a 6-in. beam port, graphite thermal column, andmore » in-core sample facility. The reactor, housed in an open 8000-gal tank of water, is designed so that the core is readily accessible. Both the control console and the peripheral counting equipment used for student projects and laboratory exercises are located in the reactor room. This arrangement provides convenience and flexibility in using the reactor for foil activations in neutron flux measurements, diffusion measurements, radioactive decay measurements, and the neutron activation of samples for analysis. In 1988, the reactor was successfully converted to low-enriched uranium fuel.« less

Control console replacement at the WPI Reactor. [Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-12-31

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. Themore » objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.« less

Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

NASA Astrophysics Data System (ADS)

Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

2018-02-01

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

The effect of mixing on fermentation of primary solids, glycerol, and biodiesel waste.

PubMed

Ghasemi, Marzieh; Randall, Andrew A

2018-03-01

In this study, the effect of mixing on volatile fatty acid (VFA) production and composition was investigated through running five identical bench-scale reactors that were filled with primary solid and dosed with either pure glycerol or biodiesel waste. Experimental results revealed that there was an inverse correlation between the mixing intensity and the VFA production. The total VFA production in the un-mixed reactor was 9,787 ± 3,601 mg COD/L, whereas in the reactor mixed at 100 rpm this dropped to 3,927 ± 1,175 mg COD/L, while both types of reactor were dosed with pure glycerol at the beginning of each cycle to reach the initial concentration of 1,000 mg/L (1,217 mg COD/L). Propionic acid was the dominant VFA in all the reactors except the reactor mixed at 30 rpm. It is hypothesized that low mixing facilitated hydrogen transfer between obligate hydrogen producing acetogens (OHPA) and hydrogen consuming acidogens in these non-methanogenic reactors. Also, in a narrower range of mixing (0 or 7 rpm), the total VFA production in biodiesel waste-fed reactors was considerably higher than that of pure glycerol-fed reactors.

Application of Reactor Antineutrinos: Neutrinos for Peace

NASA Astrophysics Data System (ADS)

Suekane, F.

2013-02-01

In nuclear reactors, 239Pu are produced along with burn-up of nuclear fuel. 239Pu is subject of safeguard controls since it is an explosive component of nuclear weapon. International Atomic Energy Agency (IAEA) is watching undeclared operation of reactors to prevent illegal production and removal of 239Pu. In operating reactors, a huge numbers of anti electron neutrinos (ν) are produced. Neutrino flux is approximately proportional to the operating power of reactor in short term and long term decrease of the neutrino flux per thermal power is proportional to the amount of 239Pu produced. Thus rector ν's carry direct and real time information useful for the safeguard purposes. Since ν can not be hidden, it could be an ideal medium to monitor the reactor operation. IAEA seeks for novel technologies which enhance their ability and reactor neutrino monitoring is listed as one of such candidates. Currently neutrino physicists are performing R&D of small reactor neutrino detectors to use specifically for the safeguard use in response to the IAEA interest. In this proceedings of the neutrino2012 conference, possibilities of such reactor neutrinos application and current world-wide R&D status are described.

Applicability of 100kWe-class of space reactor power systems to NASA manned space station missions

NASA Technical Reports Server (NTRS)

Silverman, S. W.; Willenberg, H. J.; Robertson, C.

1985-01-01

An assessment is made of a manned space station operating with sufficiently high power demands to require a multihundred kilowatt range electrical power system. The nuclear reactor is a competitor for supplying this power level. Load levels were selected at 150kWe and 300kWe. Interactions among the reactor electrical power system, the manned space station, the space transportation system, and the mission were evaluated. The reactor shield and the conversion equipment were assumed to be in different positions with respect to the station; on board, tethered, and on a free flyer platform. Mission analyses showed that the free flyer concept resulted in unacceptable costs and technical problems. The tethered reactor providing power to an electrolyzer for regenerative fuel cells on the space station, results in a minimum weight shield and can be designed to release the reactor power section so that it moves to a high altitude orbit where the decay period is at least 300 years. Placing the reactor on the station, on a structural boom is an attractive design, but heavier than the long tethered reactor design because of the shield weight for manned activity near the reactor.

Nitrate removal with lateral flow sulphur autotrophic denitrification reactor.

PubMed

Lv, Xiaomei; Shao, Mingfei; Li, Ji; Xie, Chuanbo

2014-01-01

An innovative lateral flow sulphur autotrophic denitrification (LFSAD) reactor was developed in this study; the treatment performance was evaluated and compared with traditional sulphur/limestone autotrophic denitrification (SLAD) reactor. Results showed that nitrite accumulation in the LFSAD reactor was less than 1.0 mg/L during the whole operation. Denitrification rate increased with the increased initial alkalinity and was approaching saturation when initial alkalinity exceeded 2.5 times the theoretical value. Higher influent nitrate concentration could facilitate nitrate removal capacity. In addition, denitrification efficiency could be promoted under an appropriate reflux ratio, and the highest nitrate removal percentage was achieved under reflux ratio of 200%, increased by 23.8% than that without reflux. Running resistance was only about 1/9 of that in SLAD reactor with equal amount of nitrate removed, which was the prominent excellence of the new reactor. In short, this study indicated that the developed reactor was feasible for nitrate removal from waters with lower concentrations, including contaminated surface water, groundwater or secondary effluent of municipal wastewater treatment with fairly low running resistance. The innovation in reactor design in this study may bring forth new ideas of reactor development of sulphur autotrophic denitrification for nitrate-contaminated water treatment.

76 FR 79229 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-21

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on January 19-20, 2012, 11545 Rockville... Cooling Systems for Light- Water Nuclear Power Reactors'' (Open)--The Committee will hear presentations by...

76 FR 68514 - Request for a License To Export Reactor Components

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-04

... NUCLEAR REGULATORY COMMISSION Request for a License To Export Reactor Components Pursuant to 10.../docket Number Westinghouse Electric Company Complete reactor 12 Perform seismic China. LLC, August 18... qualification equipment. of AP1000 (design) nuclear reactors. For the Nuclear Regulatory Commission. Dated this...

77 FR 16270 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-20

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0070] Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft..., ``Updated Aging Management Criteria for PWR Reactor Vessel Internal Components.'' This draft LR-ISG revises...

77 FR 23513 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-19

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0070] Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft...-ISG), LR-ISG-2011-04, ``Updated Aging Management Criteria for PWR Reactor Vessel Internal Components...

77 FR 37074 - License Amendment Request From the Alan J. Blotcky Reactor Facility

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-20

... the Alan J. Blotcky Reactor Facility AGENCY: Nuclear Regulatory Commission. ACTION: Notice of... section of this document. FOR FURTHER INFORMATION CONTACT: Theodore Smith, Project Manager, Reactor... provided the first time that a document is referenced. The Alan J. Blotcky Reactor Facility Decommissioning...

NUCLEAR REACTOR

DOEpatents

Treshow, M.

1961-09-01

A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

77 FR 8902 - Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-15

... Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide... draft regulatory guide (DG) DG-1271 ``Decommissioning of Nuclear Power Reactors.'' This guide describes... Regulatory Guide 1.184, ``Decommissioning of Nuclear Power Reactors,'' dated July 2000. This proposed...

10 CFR 2.108 - Denial of application for failure to supply information.

Code of Federal Regulations, 2012 CFR

2012-01-01

... supply information. (a) The Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and Safeguards, as appropriate, may deny an... of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear...

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear... requirements for immediate notification of the NRC by licensed operating nuclear power reactors are contained...

10 CFR 50.36 - Technical specifications.

Code of Federal Regulations, 2013 CFR

2013-01-01

... nuclear reactors are limits upon important process variables that are found to be necessary to reasonably... Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b... for nuclear reactors are settings for automatic protective devices related to those variables having...

10 CFR 2.108 - Denial of application for failure to supply information.

Code of Federal Regulations, 2011 CFR

2011-01-01

... supply information. (a) The Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and Safeguards, as appropriate, may deny an... of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear...

10 CFR 50.36 - Technical specifications.

Code of Federal Regulations, 2014 CFR

2014-01-01

... nuclear reactors are limits upon important process variables that are found to be necessary to reasonably... Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b... for nuclear reactors are settings for automatic protective devices related to those variables having...

10 CFR 50.36 - Technical specifications.

Code of Federal Regulations, 2012 CFR

2012-01-01

... nuclear reactors are limits upon important process variables that are found to be necessary to reasonably... Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b... for nuclear reactors are settings for automatic protective devices related to those variables having...

Radial blanket assembly orificing arrangement

DOEpatents

Patterson, J.F.

1975-07-01

A nuclear reactor core for a liquid metal cooled fast breeder reactor is described in which means are provided for increasing the coolant flow through the reactor fuel assemblies as the reactor ages by varying the coolant flow rate with the changing coolant requirements during the core operating lifetime. (auth)

Autonomous Control of Space Nuclear Reactors

NASA Technical Reports Server (NTRS)

Merk, John

2013-01-01

Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation. Long-duration surface missions necessitate reliable autonomous operation, and manned missions impose added requirements for failsafe reactor protection. There is a need for an advanced instrumentation and control system for space-nuclear reactors that addresses both aspects of autonomous operation and safety. The Reactor Instrumentation and Control System (RICS) consists of two functionally independent systems: the Reactor Protection System (RPS) and the Supervision and Control System (SCS). Through these two systems, the RICS both supervises and controls a nuclear reactor during normal operational states, as well as monitors the operation of the reactor and, upon sensing a system anomaly, automatically takes the appropriate actions to prevent an unsafe or potentially unsafe condition from occurring. The RPS encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device output terminals. The SCS contains a comprehensive data acquisition system to measure continuously different groups of variables consisting of primary measurement elements, transmitters, or conditioning modules. These reactor control variables can be categorized into two groups: those directly related to the behavior of the core (known as nuclear variables) and those related to secondary systems (known as process variables). Reliable closed-loop reactor control is achieved by processing the acquired variables and actuating the appropriate device drivers to maintain the reactor in a safe operating state. The SCS must prevent a deviation from the reactor nominal conditions by managing limitation functions in order to avoid RPS actions. The RICS has four identical redundancies that comply with physical separation, electrical isolation, and functional independence. This architecture complies with the safety requirements of a nuclear reactor and provides high availability to the host system. The RICS is intended to interface with a host computer (the computer of the spacecraft where the reactor is mounted). The RICS leverages the safety features inherent in Earth-based reactors and also integrates the wide range neutron detector (WRND). A neutron detector provides the input that allows the RICS to do its job. The RICS is based on proven technology currently in use at a nuclear research facility. In its most basic form, the RICS is a ruggedized, compact data-acquisition and control system that could be adapted to support a wide variety of harsh environments. As such, the RICS could be a useful instrument outside the scope of a nuclear reactor, including military applications where failsafe data acquisition and control is required with stringent size, weight, and power constraints.

Simulating Asymmetric Top Impurities in Superfluid Clusters: A para-Water Dopant in para-Hydrogen.

PubMed

Zeng, Tao; Li, Hui; Roy, Pierre-Nicholas

2013-01-03

We present the first simulation study of bosonic clusters doped with an asymmetric top molecule. The path-integral Monte Carlo method with the latest methodological advance in treating rigid-body rotation [Noya, E. G.; Vega, C.; McBride, C. J. Chem. Phys.2011, 134, 054117] is employed to study a para-water impurity in para-hydrogen clusters with up to 20 para-hydrogen molecules. The growth pattern of the doped clusters is similar in nature to that of pure clusters. The para-water molecule appears to rotate freely in the cluster. The presence of para-water substantially quenches the superfluid response of para-hydrogen with respect to the space-fixed frame.

Tower Shielding Reactor II design and operation report: Vol. 2. Safety Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holland, L. B.; Kolb, J. O.

1970-01-01

Information on the Tower Shielding Reactor II is contained in the TSR-II Design and Operation Report and in the Tower Shielding Facility Manual. The TSR-II Design and Operating Report consists of three volumes. Volume 1 is Descriptions of the Tower Shielding Reactor II and Facility; Volume 2 is Safety analysis of the Tower Shielding Reactor II; and Volume 3 is the Assembly and Testing of the Tower Shielding Reactor II Control Mechanism Housing.

Thermomechanical analysis of fast-burst reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, J.D.

1994-08-01

Fast-burst reactors are designed to provide intense, short-duration pulses of neutrons. The fission reaction also produces extreme time-dependent heating of the nuclear fuel. An existing transient-dynamic finite element code was modified specifically to compute the time-dependent stresses and displacements due to thermal shock loads of reactors. Thermomechanical analysis was then applied to determine structural feasibility of various concepts for an EDNA-type reactor and to optimize the mechanical design of the new SPR III-M reactor.

PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP

DOEpatents

Puechl, K.H.

1963-09-24

A pressurized water reactor is described having a core containing Pu/sup 240/ in which the effective microscopic neutronabsorption cross section of Pu/sup 240/ in unconverted condition decreases as the time of operation of the reactor increases, in order to compensate for loss of reactivity resulting from fission product buildup during reactor operation. This means serves to improve the efficiency of the reactor operation by reducing power losses resulting from control rods and burnable poisons. (AEC)

Seed and blanket fuel arrangement for dual-phase nuclear reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Congdon, S.P.; Fawcett, R.M.

1992-09-22

This patent describes a fuel management method for a dual-phase nuclear reactor, it comprises: installing a fuel bundle at a first core location accessed by coolant through a relatively small aperture, each of the bundles having a predetermined group of fuel elements; operating the reactor a first time; shutting down the reactor; reinstalling the fuel bundle at a second core location accessed by coolant through a relatively large aperture; and operating the reactor a second time.

Digital computer operation of a nuclear reactor

DOEpatents

Colley, R.W.

1982-06-29

A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

Digital computer operation of a nuclear reactor

DOEpatents

Colley, Robert W.

1984-01-01

A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

A Single-Granule-Level Approach Reveals Ecological Heterogeneity in an Upflow Anaerobic Sludge Blanket Reactor

PubMed Central

Mei, Ran; Narihiro, Takashi; Bocher, Benjamin T. W.; Yamaguchi, Takashi; Liu, Wen-Tso

2016-01-01

Upflow anaerobic sludge blanket (UASB) reactor has served as an effective process to treat industrial wastewater such as purified terephthalic acid (PTA) wastewater. For optimal UASB performance, balanced ecological interactions between syntrophs, methanogens, and fermenters are critical. However, much of the interactions remain unclear because UASB have been studied at a “macro”-level perspective of the reactor ecosystem. In reality, such reactors are composed of a suite of granules, each forming individual micro-ecosystems treating wastewater. Thus, typical approaches may be oversimplifying the complexity of the microbial ecology and granular development. To identify critical microbial interactions at both macro- and micro- level ecosystem ecology, we perform community and network analyses on 300 PTA–degrading granules from a lab-scale UASB reactor and two full-scale reactors. Based on MiSeq-based 16S rRNA gene sequencing of individual granules, different granule-types co-exist in both full-scale reactors regardless of granule size and reactor sampling depth, suggesting that distinct microbial interactions occur in different granules throughout the reactor. In addition, we identify novel networks of syntrophic metabolic interactions in different granules, perhaps caused by distinct thermodynamic conditions. Moreover, unseen methanogenic relationships (e.g. “Candidatus Aminicenantes” and Methanosaeta) are observed in UASB reactors. In total, we discover unexpected microbial interactions in granular micro-ecosystems supporting UASB ecology and treatment through a unique single-granule level approach. PMID:27936088

Health physics aspects of advanced reactor licensing reviews

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hinson, C.S.

1995-03-01

The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on {open_quotes}next-generation{close_quotes} reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovativemore » design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs currently being reviewed by the NRC.« less

Numerical study of the effects of lamp configuration and reactor wall roughness in an open channel water disinfection UV reactor.

PubMed

Sultan, Tipu

2016-07-01

This article describes the assessment of a numerical procedure used to determine the UV lamp configuration and surface roughness effects on an open channel water disinfection UV reactor. The performance of the open channel water disinfection UV reactor was numerically analyzed on the basis of the performance indictor reduction equivalent dose (RED). The RED values were calculated as a function of the Reynolds number to monitor the performance. The flow through the open channel UV reactor was modelled using a k-ε model with scalable wall function, a discrete ordinate (DO) model for fluence rate calculation, a volume of fluid (VOF) model to locate the unknown free surface, a discrete phase model (DPM) to track the pathogen transport, and a modified law of the wall to incorporate the reactor wall roughness effects. The performance analysis was carried out using commercial CFD software (ANSYS Fluent 15.0). Four case studies were analyzed based on open channel UV reactor type (horizontal and vertical) and lamp configuration (parallel and staggered). The results show that lamp configuration can play an important role in the performance of an open channel water disinfection UV reactor. The effects of the reactor wall roughness were Reynolds number dependent. The proposed methodology is useful for performance optimization of an open channel water disinfection UV reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

PARTIAL ECONOMIC STUDY OF STEAM COOLED HEAVY WATER MODERATED REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1960-04-01

Steam-cooled reactors are compared with CAHDU for costs of Calandria tubes, pressure tubes. heavy water moderator, heavy water reflector, fuel supply, heat exchanger, and turbine generator. A direct-cycle lightsteam-cooled heavy- water-moderated pressure-tube reactor formed the basic reactor design for the study. Two methods of steam circulation through the reactor were examined. In both cases the steam was generated outside the reactor and superheated in the reactor core. One method consisted of a series of reactor and steam generator passes. The second method consisted of the Loeffler cycle and its modifications. The fuel was assumed to be natural cylindrical UO/sub 2/more » pellets sheathed in a hypothetical material with the nuclear properties of Zircaloy, but able to function at temperatures to 900 deg F. For the conditions assumed, the longer the rod, the higher the outlet temperature and therefore the higher the efficiency. The turbine cycle efficiency was calculated on the assumption that suitable steam generators are available. As the neutron losses to the pressure tubes were significant, an economic analysis of insulated pressure tubes is included. A description of the physics program for steam-cooled reactors is included. Results indicated that power from the steam-cooled reactor would cost 1.4 mills/ kwh compared with 1.25 mills/kwh for CANDU. (M.C.G.)« less

Dismantling of Loop-Type Channel Equipment of MR Reactor in NRC 'Kurchatov Institute' - 13040

DOE Office of Scientific and Technical Information (OSTI.GOV)

Volkov, Victor; Danilovich, Alexey; Zverkov, Yuri

2013-07-01

In 2009 the project of decommissioning of MR and RTF reactors was developed and approved by the Expert Authority of the Russian Federation (Gosexpertiza). The main objective of the decommissioning works identified in this project: - complete dismantling of reactor equipment and systems; - decontamination of reactor premises and site in accordance with the established sanitary and hygienic standards. At the preparatory stage (2008-2010) of the project the following works were executed: loop-type channels' dismantling in the storage pool; experimental fuel assemblies' removal from spent fuel repositories in the central hall; spent fuel assembly removal from the liquid-metal-cooled loop-type channelmore » of the reactor core and its placement into the SNF repository; and reconstruction of engineering support systems to the extent necessary for reactor decommissioning. The project assumes three main phases of dismantling and decontamination: - dismantling of equipment/pipelines of cooling circuits and loop-type channels, and auxiliary reactor equipment (2011-2012); - dismantling of equipment in underground reactor premises and of both MR and RTF in-vessel devices (2013-2014); - decontamination of reactor premises; rehabilitation of the reactor site; final radiation survey of reactor premises, loop-type channels and site; and issuance of the regulatory authorities' de-registration statement (2015). In 2011 the decommissioning license for the two reactors was received and direct MR decommissioning activities started. MR primary pipelines and loop-type facilities situated in the underground reactor hall were dismantled. Works were also launched to dismantle the loop-type channels' equipment in underground reactor premises; reactor buildings were reconstructed to allow removal of dismantled equipment; and the MR/RTF decommissioning sequence was identified. In autumn 2011 - spring 2012 results of dismantling activities performed are: - equipment from underground rooms (No. 66, 66A, 66B, 72, 64, 63) - as well as from water and gas loop corridors - was dismantled, with the total radwaste weight of 53 tons and the total removed activity of 5,0 x 10{sup 10} Bq; - loop-type channel equipment from underground reactor hall premises was dismantled; - 93 loop-type channels were characterized, chopped and removed, with radwaste of 2.6 x 10{sup 13} Bq ({sup 60}Co) and 1.5 x 10{sup 13} Bq ({sup 137}Cs) total activity removed from the reactor pool, fragmented and packaged. Some of this waste was placed into the high-level waste (HLW) repository of the Center. Dismantling works were executed with application of remotely operated mechanisms, which promoted decrease of radiation impact on the personnel. The average individual dose for the personnel was 1.9 mSv/year in 2011, and the collective dose is estimated as 0.0605 man x Sv/year. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

Progress is reported on fundamental research in: crystal physics, reactions at metal surfaces, spectroscopy of ionic media, structure of metals, theory of alloying, physical properties, sintering, deformation of crystalline solids, x ray diffraction, metallurgy of superconducting materials, and electron microscope studies. Long-randge applied research studies were conducted for: zirconium metallurgy, materials compatibility, solid reactions, fuel element development, mechanical properties, non-destructive testing, and high-temperature materials. Reactor development support work was carried out for: gas-cooled reactor program, molten-salt reactor, high-flux isotope reactor, space-power program, thorium-utilization program, advanced-test reactor, Army Package Power Reactor, Enrico Fermi fast-breeder reactor, and water desalination program. Other programmore » activities, for which research was conducted, included: thermonuclear project, transuraniunn program, and post-irradiation examination laboratory. Separate abstracts were prepared for 30 sections of the report. (B.O.G.)« less

Analysis of the SL-1 Accident Using RELAPS5-3D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francisco, A.D. and Tomlinson, E. T.

2007-11-08

On January 3, 1961, at the National Reactor Testing Station, in Idaho Falls, Idaho, the Stationary Low Power Reactor No. 1 (SL-1) experienced a major nuclear excursion, killing three people, and destroying the reactor core. The SL-1 reactor, a 3 MW{sub t} boiling water reactor, was shut down and undergoing routine maintenance work at the time. This paper presents an analysis of the SL-1 reactor excursion using the RELAP5-3D thermal-hydraulic and nuclear analysis code, with the intent of simulating the accident from the point of reactivity insertion to destruction and vaporization of the fuel. Results are presented, along with amore » discussion of sensitivity to some reactor and transient parameters (many of the details are only known with a high level of uncertainty).« less

Axi-symmetrical flow reactor for [sup 196]Hg photochemical enrichment

DOEpatents

Grossman, M.W.

1991-04-30

The present invention is directed to an improved photochemical reactor useful for the isotopic enrichment of a predetermined isotope of mercury, especially, [sup 196]Hg. Specifically, two axi-symmetrical flow reactors were constructed according to the teachings of the present invention. These reactors improve the mixing of the reactants during the photochemical enrichment process, affording higher yields of the desired [sup 196]Hg product. Measurements of the variation of yield (Y) and enrichment factor (E) along the flow axis of these reactors indicates very substantial improvement in process uniformity compared to previously used photochemical reactor systems. In one preferred embodiment of the present invention, the photoreactor system was built such that the reactor chamber was removable from the system without disturbing the location of either the photochemical lamp or the filter employed therewith. 10 figures.

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

NASA Astrophysics Data System (ADS)

Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-01

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bahri, Che Nor Aniza Che Zainul, E-mail: anizazainul@gmail.com; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-29

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclearmore » waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.« less

Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forsberg, C.W.; Reich, W.J.

1991-09-01

The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactormore » concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.« less

Locating hot and cold-legs in a nuclear powered steam generation system

DOEpatents

Ekeroth, D.E.; Corletti, M.M.

1993-11-16

A nuclear reactor steam generator includes a reactor vessel for heating water and a steam generator with a pump casing at the lowest point on the steam generator. A cold-leg pipe extends horizontally between the steam generator and the reactor vessel to return water from the steam generator to the reactor vessel. The bottom of the cold-leg pipe is at a first height above the bottom of the reactor vessel. A hot-leg pipe with one end connected to the steam generator and a second end connected to the reactor vessel has a first pipe region extending downwardly from the steam generator to a location between the steam generator and the reactor vessel at which a bottom of the hot-leg pipe is at a second height above the bottom of the reactor vessel. A second region extends from that location in a horizontal direction at the second height to the point at which the hot-leg pipe connects to the reactor vessel. A pump is attached to the casing at a location below the first and second heights and returns water from the steam generator to the reactor vessel over the cold-leg. The first height is greater than the second height and the bottom of the steam generator is at a height above the bottom of the reactor vessel that is greater than the first and second heights. A residual heat recovery pump is below the hot-leg and has an inlet line from the hot-leg that slopes down continuously to the pump inlet. 2 figures.

Locating hot and cold-legs in a nuclear powered steam generation system

DOEpatents

Ekeroth, Douglas E.; Corletti, Michael M.

1993-01-01

A nuclear reactor steam generator includes a reactor vessel for heating water and a steam generator with a pump casing at the lowest point on the steam generator. A cold-leg pipe extends horizontally between the steam generator and the reactor vessel to return water from the steam generator to the reactor vessel. The bottom of the cold-leg pipe is at a first height above the bottom of the reactor vessel. A hot-leg pipe with one end connected to the steam generator and a second end connected to the reactor vessel has a first pipe region extending downwardly from the steam generator to a location between the steam generator and the reactor vessel at which a bottom of the hot-leg pipe is at a second height above the bottom of the reactor vessel. A second region extends from that location in a horizontal direction at the second height to the point at which the hot-leg pipe connects to the reactor vessel. A pump is attached to the casing at a location below the first and second heights and returns water from the steam generator to the reactor vessel over the cold-leg. The first height is greater than the second height and the bottom of the steam generator is at a height above the bottom of the reactor vessel that is greater than the first and second heights. A residual heat recovery pump is below the hot-leg and has an inlet line from the hot-leg that slopes down continuously to the pump inlet.

77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-12

... for Passive Advanced Light Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... Passive Advanced Light Water Reactors.'' The current SRP does not contain guidance on the proposed RTNSS for Passive Advance Light Water Reactors. DATES: Submit comments by November 13, 2012. Comments...

Animal Guts as Ideal Reactors: An Open-Ended Project for a Course in Kinetics and Reactor Design.

ERIC Educational Resources Information Center

Carlson, Eric D.; Gast, Alice P.

1998-01-01

Presents an open-ended project tailored for a senior kinetics and reactor design course in which basic reactor design equations are used to model the digestive systems of several animals. Describes the assignment as well as the results. (DDR)

77 FR 69900 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-21

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on December 6-8, 2012, 11545 Rockville... Recommendations (SECY-12-0064), (3) Venting Systems for Boiling Water Reactors (BWRs) with Mark I and Mark II...

78 FR 57904 - Request for a License To Export; Reactor Components

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-20

... NUCLEAR REGULATORY COMMISSION Request for a License To Export; Reactor Components Pursuant to 10..., systems, related reactors. operation of AP- XR177, 11006121. equipment, and 1000 (design) spare parts. nuclear reactors. Dated this 16th day of September 2013 in Rockville, Maryland. For The Nuclear Regulatory...

9 CFR 78.31 - Brucellosis reactor swine.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 9 Animals and Animal Products 1 2011-01-01 2011-01-01 false Brucellosis reactor swine. 78.31... Restrictions on Interstate Movement of Swine Because of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for immediate slaughter as follows: (1...

76 FR 57082 - Advisory Committee on Reactor Safeguards; Meeting of the ACRS Subcommittee on Materials...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-15

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Meeting of the ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels Revision to September 21, 2011, ACRS Meeting; Federal... Reactor Fuels is being revised to correct the meeting date to Wednesday, September 21, 2011. The notice of...

77 FR 64563 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-22

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on November 1-3, 2012, 11545 Rockville...-Term Core Cooling Approach for the Advanced Boiling Water Reactor (ABWR) Design for South Texas Project...

10 CFR 54.25 - Report of the Advisory Committee on Reactor Safeguards.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Report of the Advisory Committee on Reactor Safeguards. 54.25 Section 54.25 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REQUIREMENTS FOR RENEWAL OF... Reactor Safeguards. Each renewal application will be referred to the Advisory Committee on Reactor...

9 CFR 78.22 - Brucellosis reactor bison.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 9 Animals and Animal Products 1 2011-01-01 2011-01-01 false Brucellosis reactor bison. 78.22... Restrictions on Interstate Movement of Bison Because of Brucellosis § 78.22 Brucellosis reactor bison. (a) Destination. Brucellosis reactor bison may be moved interstate only for immediate slaughter as follows: (1...

76 FR 76442 - Advisory Committee On Reactor Safeguards Meeting of The ACRS Subcommittee on Materials...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-07

... NUCLEAR REGULATORY COMMISSION Advisory Committee On Reactor Safeguards Meeting of The ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels Revision to December 15, 2011, ACRS Meeting Federal... & Reactor Fuels scheduled to be held on December 15, 2011, is being revised to notify the following: The...

9 CFR 78.31 - Brucellosis reactor swine.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Brucellosis reactor swine. 78.31... Restrictions on Interstate Movement of Swine Because of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for immediate slaughter as follows: (1...

10 CFR 100.10 - Factors to be considered when evaluating sites.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Section 100.10 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA Evaluation Factors for Stationary Power Reactor Site Applications Before January 10, 1997 and for Testing Reactors § 100... include those relating both to the proposed reactor design and the characteristics peculiar to the site...

9 CFR 78.7 - Brucellosis reactor cattle.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 9 Animals and Animal Products 1 2011-01-01 2011-01-01 false Brucellosis reactor cattle. 78.7... Restrictions on Interstate Movement of Cattle Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a) Destination. Brucellosis reactor cattle may be moved interstate only for immediate slaughter as follows: (1...

9 CFR 78.7 - Brucellosis reactor cattle.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Brucellosis reactor cattle. 78.7... Restrictions on Interstate Movement of Cattle Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a) Destination. Brucellosis reactor cattle may be moved interstate only for immediate slaughter as follows: (1...

10 CFR 54.25 - Report of the Advisory Committee on Reactor Safeguards.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Report of the Advisory Committee on Reactor Safeguards. 54.25 Section 54.25 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REQUIREMENTS FOR RENEWAL OF... Reactor Safeguards. Each renewal application will be referred to the Advisory Committee on Reactor...

10 CFR 2.1115 - Designation of issues for adjudicatory hearing.

Code of Federal Regulations, 2010 CFR

2010-01-01

... at Civilian Nuclear Power Reactors § 2.1115 Designation of issues for adjudicatory hearing. (a) After... reactor already licensed to operate at the site, or any civilian nuclear power reactor for which a... the issuance of a construction permit or operating license for a civilian nuclear power reactor at...

9 CFR 78.22 - Brucellosis reactor bison.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Brucellosis reactor bison. 78.22... Restrictions on Interstate Movement of Bison Because of Brucellosis § 78.22 Brucellosis reactor bison. (a) Destination. Brucellosis reactor bison may be moved interstate only for immediate slaughter as follows: (1...

10 CFR 100.10 - Factors to be considered when evaluating sites.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Section 100.10 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA Evaluation Factors for Stationary Power Reactor Site Applications Before January 10, 1997 and for Testing Reactors § 100... include those relating both to the proposed reactor design and the characteristics peculiar to the site...

78 FR 56174 - In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-12

...-core thermocouples at different elevations and radial positions throughout the reactor core to enable... different elevations and radial positions throughout the reactor core to enable NPP operators to accurately... NPPs with in-core thermocouples at different elevations and radial positions throughout the reactor...

FAST NEUTRON REACTOR

DOEpatents

Soodak, H.; Wigner, E.P.

1961-07-25

A reactor comprising fissionable material in concentration sufficiently high so that the average neutron enengy within the reactor is at least 25,000 ev is described. A natural uranium blanket surrounds the reactor, and a moderating reflector surrounds the blanket. The blanket is thick enough to substantially eliminate flow of neutrons from the reflector.

75 FR 70042 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-16

... NUCLEAR REGULATORY COMMISSION [Docket Nos (Redacted), License Nos (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I. The licensees identified in...

75 FR 79423 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-20

... NUCLEAR REGULATORY COMMISSION [Docket Nos. (Redacted), License Nos.: (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I The licensees identified in...

Determination of parameters of a nuclear reactor through noise measurements

DOEpatents

Cohn, C.E.

1975-07-15

A method of measuring parameters of a nuclear reactor by noise measurements is described. Noise signals are developed by the detectors placed in the reactor core. The polarity coincidence between the noise signals is used to develop quantities from which various parameters of the reactor can be calculated. (auth)

CONTROL OF VOLATILE ORGANIC COMPOUNDS BY AN AC ENERGIZED FERROELECTRIC PELLET REACTOR AND A PULSED CORONA REACTOR

EPA Science Inventory

The paper gives results of a study to develop baseline engineering data to demonstrate the feasibility of application of plasma reactors to the destruction of various volatile organic compounds at ppm levels. Two laboratory-scale reactors, an alternating current energized ferroel...

157. ARAIII Reactor building (ARA608) Main gas loop mechanical flow ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

157. ARA-III Reactor building (ARA-608) Main gas loop mechanical flow sheet. This drawing was selected as a typical example of mechanical arrangements within reactor building. Aerojet-general 880-area/GCRE-0608-50-013-102634. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR

DOEpatents

Tokarz, R.D.

1981-10-27

This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

78 FR 71675 - Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor Correspondence

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-29

... correspondence to addressees and subscribers through a computer-based email distribution system. Since then, the... Electronic Operating Reactor Correspondence The U.S. Nuclear Regulatory Commission (NRC) is issuing this... available operating reactor licensing correspondence, effective December 9, 2013. Official agency records...

9 CFR 78.31 - Brucellosis reactor swine.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 9 Animals and Animal Products 1 2012-01-01 2012-01-01 false Brucellosis reactor swine. 78.31... Restrictions on Interstate Movement of Swine Because of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for immediate slaughter as follows: (1...

9 CFR 78.31 - Brucellosis reactor swine.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 9 Animals and Animal Products 1 2013-01-01 2013-01-01 false Brucellosis reactor swine. 78.31... Restrictions on Interstate Movement of Swine Because of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for immediate slaughter as follows: (1...

9 CFR 78.31 - Brucellosis reactor swine.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 9 Animals and Animal Products 1 2014-01-01 2014-01-01 false Brucellosis reactor swine. 78.31... Restrictions on Interstate Movement of Swine Because of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for immediate slaughter as follows: (1...

Utilization of the Recycle Reactor in Determining Kinetics of Gas-Solid Catalytic Reactions.

ERIC Educational Resources Information Center

Paspek, Stephen C.; And Others

1980-01-01

Describes a laboratory scale reactor that determines the kinetics of a gas-solid catalytic reaction. The external recycle reactor construction is detailed with accompanying diagrams. Experimental details, application of the reactor to CO oxidation kinetics, interphase gradients, and intraphase gradients are discussed. (CS)

A Semi-Batch Reactor Experiment for the Undergraduate Laboratory

ERIC Educational Resources Information Center

Derevjanik, Mario; Badri, Solmaz; Barat, Robert

2011-01-01

This experiment and analysis offer an economic yet challenging semi-batch reactor experience. Household bleach is pumped at a controlled rate into a batch reactor containing pharmaceutical hydrogen peroxide solution. Batch temperature, product molecular oxygen, and the overall change in solution conductivity are metered. The reactor simulation…

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...