REMOCIÓN DE ARSÉNICO (V ASISTIDA POR OXIDACIÓN UV SOLAR EN UN FOTO-REACTOR TUBULAR DE SECCIÓN CIRCULAR

Directory of Open Access Journals (Sweden)

Ramiro Escalera Vásquez

2010-01-01

Full Text Available Se ha construido y caracterizado un foto-reactor tubular de sección circular para su aplicación al tratamiento de aguas subterráneas contaminadas con Arsénico, As(V, utilizando las técnica de la Remoción de Arsénico por Oxidación Solar (RAOS. El concentrador solar que posee una capacidad de radiación equivalente a 2,8 soles, fue construido reciclando materiales desechados: tubos de vidrio proveniente de lámparas de Ne y tubos de desagüe sanitario de 6” (PVC, recubiertos por láminas de aluminio. Pruebas simultáneas sin agitación,realizadas aplicando la radiación UV solar a aguas sintéticas, demostraron que la remoción de As(V en el foto-reactor es más rápida queen un tubo de vidrio sólo y en una botella PET de 2 litros, logrando remociones mayores al 98% en todos los casos. Los tiempos para la aparición de los flóculos de complejo Fe-citrato fueron de 40, 50 y 90 min respectivamente, para intensidades de radiación UVA integral (290-390 nm entre 50 y 70 Wm-2. Pruebas de irradiación seguidas de agitación controlada a 30-33 s-1 de gradiente de velocidad, demostraron que el foto-reactor acelera el proceso de formación de flóculos fácilmente sedimentables al cabo de 20-30 min de agitación. Los tiempos de irradiación óptimos para el foto-reactor, el tubo y la botella son de 15, 25 y 60 min, respectivamente. Pruebas en régimen de flujo continuo en un foto-reactor de aproximadamente 1 m2 de área, con un tiempo de residencia hidráulica (igual al tiempo de irradiación de 15 min, mostraron la formación inmediata de flóculos fácilmente sedimentables cuando se agitan a 33 s-1 durante 20-30 min, lográndose una remoción del 98,36% una concentración remanente de 16,5 mgL-1 de As(V en aguas decantadas. Esto significa que se pueden tratar aproximadamente 130 Lm-2 en una jornada de 6 horas de radiación UVA de 50-70 Wm-2 de intensidad.

Design of a reactor core in the Oma Full MOX-ABWR

International Nuclear Information System (INIS)

Hama, Teruo

1999-01-01

The Electric Power Development Co., Ltd. has progressed a construction plan on an improved boiling-water reactor aiming at loading of MOX fuel in all reactor cores (full MOX-ABWR) at Oma-cho, Aomori prefecture, which is a last stage on application of approval on establishment at present. Here were described on outlines of reactor core in the full MOX-ABWR and its safety evaluation. For the full MOX-ABWR loading MOX fuel assembly into all reactor core, thermal and mechanical design analysis of fuel bars and core design analysis were conducted. As a result, it was confirmed that judgement standards in mixed core of MOX fuel and uranium fuel were also applicable as well as that in uranium fuel. (G.K.)

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

International Nuclear Information System (INIS)

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

1995-06-01

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

Performance evaluation of full scale UASB reactor in treating stillage wastewater

Directory of Open Access Journals (Sweden)

A.Mirsepasi , H. R. Honary , A. R. Mesdaghinia, A. H. Mahvi , H. Vahid , H. Karyab

2006-04-01

Full Text Available Upflow anaerobic sludge blanket (UASB reactors have been widely used for treatment of industrial wastewater. In this study two full-scale UASB reactors were investigated. Volume of each reactor was 420 m3. Conventional parameters such as pH, temperature and efficiency of COD, BOD, TOC removal in each reactor were investigated. Also several initial parameters in designing and operating of UASB reactors, such as upflow velocity, organic loading rate (OLR and hydraulic retention time were investigated. After modifying in operation conditions in UASB-2 reactor, average COD removal efficiency at OLR of 10–11 kg COD / m3 day was 55 percent. In order to prevent solids from settling, upflow velocity was increased to 0.35 m/h. Also to prevent solids from settling, the hydraulic retention time of wastewater in UASB-2 reactor was increased from 200 to 20 hours. This was expected that with good operation of UASB-2 reactor and with expanding of granules in the bed of the reactor, COD removal efficiency will be increased to more than 80 percent. But, because of deficiency on granulation and operation in UASB-2 reactor, this was not achieved. COD removal efficiency in the UASB-1 reactor was little. To enhance COD efficiency of UASB-1 reactor, several parameters were needed to be changed. These changes included enhancing of OLRs and upflow velocity, decreasing hydraulic retention time and operating with new sludge.

Full-Scale Continuous Mini-Reactor Setup for Heterogeneous Grignard Alkylation of a Pharmaceutical Intermediate

DEFF Research Database (Denmark)

Pedersen, Michael Jønch; Holm, Thomas; Rahbek, Jesper P.

2013-01-01

A reactor setup consisting of two reactors in series has been implemented for a full-scale, heterogeneous Grignard alkylation. Solutions pass from a small filter reactor into a static mixer reactor with multiple side entries, thus combining continuous stirred tank reactor (CSTR) and plug flow...

Characters of neutron noise in full-size molten salt reactor

International Nuclear Information System (INIS)

Wang, Jiangmeng; Cao, Xinrong

2015-01-01

Highlights: • The larger system size makes full-size MSR deviate from point kinetic behavior. • The increasing velocity has non-monotonic effect on the effective delayed neutron fraction. • The amplitude of Green’s function at low frequencies is inversely proportional to the effective delayed neutron fraction. • The range of plateau region is smaller due to the more prominent point kinetic effect. - Abstract: In the present paper, the frequency-dependent and space-dependent behavior of the neutron noise in a full-size Molten Salt Reactor (MSR) is investigated. The theoretical models considering the fuel circulation are established based on one-group neutron diffusion theory. Green’s function of the neutron noise induced by a propagating perturbation is introduced with linear noise theory, due to the small perturbation. The equations are numerically calculated by developing a code, in which the eigenfunction expansion method is adopted. The static results show that the effective delayed neutron fraction changes non-monotonically with the increasing fuel velocity. In the dynamic case, the results are compared to those obtained in 1D MSR and a traditional reactor, in order to figure out the effects of both the fuel circulation and the system size. It is found that there is no difference in 1D and 3D MSR systems from the view of fuel circulation, i.e., the fuel circulation enhances the spatial neutronic coupling and leads to the stronger point kinetic effect. The more prominent space-dependent effect founded in 3D traditional reactors is also observed in the MSR, due to the looser neutronic coupling and the unique singularity of Green’s function in the location of the perturbation. Another interesting finding is that Green’s function for low frequencies changes non-monotonically with increasing velocity. The largest magnitude of Green’s function is observed at the velocity where the effective delayed neutron fraction reaches its minimum. Finally, the

Design review of the N Reactor

International Nuclear Information System (INIS)

1986-09-01

This review of the design features of the N Reactor was initiated at the request of the Secretary of Energy, John S. Herrington, shortly after, and as a consequence of, reports of the accident at the Soviet reactor complex located at Chernobyl, on April 26, 1986. In the review, special attention was given to those plant systems which are most important in preventing the release of radioactive materials from the plant in the event of combined major equipment failures and human errors. Also, the review studied the potential effects of various severe accident sequences, and addressed the question of whether an event similar in causes or consequences to the Chernobyl accident could occur in the N Reactor. In light of experiences at both Three Mile Island and Chernobyl, the potential for accumulation of hydrogen in excess of flammable limits was given particular attention. The review team was also asked to identify possible improvements to the N Reactor plant, and to evaluate the effects and significance of service-induced degradation. The overall conclusion of the design review is that the N Reactor is safe to operate and that there is no reason to stop or alter its operation in any major respect at this time. Certain additional analyses and testing, are recommended to provide a firmer basis for decisions on long-term operation and on measures which may be needed in the future to accommodate long-term operation

Full-scale leaching study of commercial reactor waste forms

International Nuclear Information System (INIS)

Kalb, P.D.; Colombo, P.

1984-01-01

This paper describes a full-scale leaching experiment which has been conducted at Brookhaven National Laboratory (BNL) to study the release of radionuclides from actual commercial reactor waste forms. While many studies characterizing the leaching behavior of simulated laboratory-scale waste forms have been performed, this program represents one of the first attempts in the United States to quantify activity releases for real, full-scale waste forms. 5 references, 5 figures, 1 table

Full Core modeling techniques for research reactors with irregular geometries using Serpent and PARCS applied to the CROCUS reactor

International Nuclear Information System (INIS)

Siefman, Daniel J.; Girardin, Gaëtan; Rais, Adolfo; Pautz, Andreas; Hursin, Mathieu

2015-01-01

Highlights: • Modeling of research reactors. • Serpent and PARCS coupling. • Lattice physics codes modeling techniques. - Abstract: This paper summarizes the results of modeling methodologies developed for the zero-power (100 W) teaching and research reactor CROCUS located in the Laboratory for Reactor Physics and Systems Behavior (LRS) at the Swiss Federal Institute of Technology in Lausanne (EPFL). The study gives evidence that the Monte Carlo code Serpent can be used effectively as a lattice physics tool for small reactors. CROCUS’ core has an irregular geometry with two fuel zones of different lattice pitches. This and the reactor’s small size necessitate the use of nonstandard cross-section homogenization techniques when modeling the full core with a 3D nodal diffusion code (e.g. PARCS). The primary goal of this work is the development of these techniques for steady-state neutronics and future transient neutronics analyses of not only CROCUS, but research reactors in general. In addition, the modeling methods can provide useful insight for analyzing small modular reactor concepts based on light water technology. Static computational models of CROCUS with the codes Serpent and MCNP5 are presented and methodologies are analyzed for using Serpent and SerpentXS to prepare macroscopic homogenized group cross-sections for a pin-by-pin model of CROCUS with PARCS. The most accurate homogenization scheme lead to a difference in terms of k eff of 385 pcm between the Serpent and PARCS model, while the MCNP5 and Serpent models differed in terms of k eff by 13 pcm (within the statistical error of each simulation). Comparisons of the axial power profiles between the Serpent model as a reference and a set of PARCS models using different homogenization techniques showed a consistent root-mean-square deviation of ∼8%, indicating that the differences are not due to the homogenization technique but rather arise from the definition of the diffusion coefficients

Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

International Nuclear Information System (INIS)

Devan, K.; Bachchan, Abhitab; Riyas, A.; Sathiyasheela, T.; Mohanakrishnan, P.; Chetal, S.C.

2011-01-01

Highlights: → In this study we made physics designs of experimental metal fast reactor cores. → Aim is for full-scale demonstration of fuel assemblies in a commercial power reactor. → Minimum power with adequate safety is considered. → In addition, fuel sustainability is also considered in the design. → Sodium bonded U-Pu-6%Zr and mechanically bonded U-Pu alloys are used. - Abstract: Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.

EVALUACIÓN Y CARACTERIZACIÓN MINERALÓGICA DEL PROCESO DE BIOOXIDACIÓN EN UN REACTOR CONTINUO DE TANQUE AGITADO

Directory of Open Access Journals (Sweden)

DIANA ARROYAVE G.

2010-01-01

Full Text Available La biooxidación del mineral refractario de oro de la mina El Zancudo (TitiribíAntioquia se realizó en un reactor continuo de tanque agitado usando microorganismos nativos acidófilos compatibles con Acidithiobacillus ferrooxidans y Acidithiobacillus thiooxidans. El reactor se operó inicialmente en discontinuo para alcanzar la máxima concentración de hierro férrico en solución, antes de iniciar el proceso en continuo. La caracterización mineralógica se hizo a muestras recolectadas en discontinuo, estado transitorio y estacionario en continuo, usando Microscopia Electrónica de Barrido (SEM y Difracción de Rayos X (DRX. La caracterización mineralógica mostró una oxidación avanzada de la pirita y arsenopirita en discontinuo y parcial en continuo. Adicionalmente, se encontró la formación de silicatos, jarosita y brushita. Los resultados indican que el sistema alcanzó el estado estacionario después de 8 días de operación en continuo, logrando una concentración de hierro férrico en solución de 8.3 g/l, correspondiente a un porcentaje de extracción de oro y plata de 78 y 80 %, respectivamente.

CONSTRUCCIÓN DE UN REACTOR DISCONTINUO PARA LA OBTENCIÓN DE BIODIESEL A PARTIR DEL ACEITE DE Ricinus communis

Directory of Open Access Journals (Sweden)

Yolimar Fernández

2014-01-01

Full Text Available Se construyó un reactor discontinuo para obtener biodiesel a partir de 5 litros de extracto obtenido de la semilla de Ricinus communis. El reactor es de acero inoxidable, con longitud de 29 cm; diámetro interno de 15,24 cm y fondo cónico de 20cm de largo, espesor de la pared de 0,2cm, resistencia tubular de 1000 W y motor de 110 volt. Se extrajo y se comparó con las normas respectivas las propiedades físicas y químicas del aceite crudo. Se realizaron pruebas preliminares de transesterificación del aceite catalizadas con NaOH para constatar la viabilidad de la reacción y definir las condiciones operacionales. El biodiesel obtenido fue caracterizado y comparado con referencias presentes en la literatura. Los resultaron mostraron que es posible obtener el biocombustible en el reactor discontinuo con un grado de conversión 88%; confirmando su aplicación en reacciones de transesterificación en medio básico.

State-of-the-art incore detector system provides operational and safety benefits: Example, Hanford N Reactor

International Nuclear Information System (INIS)

Toffer, H.

1988-08-01

A presentation on the operational and safety benefits that can be derived from a state-of-the-art incore neutron monitoring system has been prepared for the DOE/ANL training course on ''The Potential Safety Impact of New and Emerging Technologies on the Operation of DOE Nuclear Facilities.'' Advanced incore neutron flux monitoring systems have been installed in some commercial reactors and should be considered for any new reactor designs or as backfits to existing plants. The recent installation of such a system at the Hanford N Reactor is used as an example in this presentation. Unfortunately, N Reactor has been placed in a cold standby condition and the full core incore system has not been tested under power conditions. Nevertheless, the evaluations that preceded the installation of the full core system provide interesting insight into the operational and safety benefits that could be expected

Validación de la limpieza del reactor empleado en la preparación de medicamentos

Directory of Open Access Journals (Sweden)

Lisseux Castilla Valentín

2001-04-01

Full Text Available En la actualidad, la validación de los procesos se ha convertido en una imperiosa necesidad de la Industria Médico-Farmacéutica, para garantizar la calidad de sus productos y lograr la comercialización de estos. En esta dirección se comenzaron los trabajos de validación de la limpieza de los equipos y se realizó la validación de la limpieza del reactor SEN, que se utiliza para la preparación de los inyectables. Para lograr este objetivo, se elaboró una metodología de trabajo, se seleccionaron los métodos analíticos más apropiados, se establecieron los criterios de aceptación y se escribió un protocolo de validación, que constituyó la herramienta fundamental de trabajo. Posteriormente se realizó la validación de la limpieza del reactor, y se concluyó que el procedimiento de limpieza, aunque no garantiza total eliminación de los residuos del producto, sí cumple con los criterios de aceptación establecidos.At present, the validation of the processes is an imperative necessity of the Medicopharmaceutical Industry to guarantee the quality of its products and to commercialize them. To this end, the cleaning of the equipment began to be validated and the validation of the cleaning of the SEN reactor that is used for the preparation of injections was carried out. To attain this goal, a working methodology was designed, the most appropiate analytical methods were selected, the acceptance criteria were established and a validation protocol was written that was the fundamental working tool. Later on, the validation of the cleaning of the reactor was made and it was concluded that although the cleaning procedure does not guarantee the total elimination of the residuals of the product, it fulfills the established acceptance criteria.

Application of assembly module to high-temperature gas-cooled reactor full-scope simulation system

International Nuclear Information System (INIS)

Li Sifeng; Li Fu; Ma Yuanle; Shi Lei

2007-01-01

According to the circumstances that exist in the reactor full-scope simulators development as long development cycle, very difficult upgrade and narrow range of applicability, a kind of new model was developed based on assembly module which root in Linux kernel and successfully applied to the design of high-temperature gas-cooled reactor full-scope simulator system. The simulation results are coincident with the experimental ones, and it indicates that the new model based on assembly module is feasible to design of high-temperature gas cooled reactor simulation system. (authors)

Analysis of core damage frequency due to external events at the DOE [Department of Energy] N-Reactor

International Nuclear Information System (INIS)

Lambright, J.A.; Bohn, M.P.; Daniel, S.L.; Baxter, J.T.; Johnson, J.J.; Ravindra, M.K.; Hashimoto, P.O.; Mraz, M.J.; Tong, W.H.; Conoscente, J.P.; Brosseau, D.A.

1990-11-01

A complete external events probabilistic risk assessment has been performed for the N-Reactor power plant, making full use of all insights gained during the past ten years' developments in risk assessment methodologies. A detailed screening analysis was performed which showed that all external events had negligible contribution to core damage frequency except fires, seismic events, and external flooding. A limited scope analysis of the external flooding risk indicated that it is not a major risk contributor. Detailed analyses of the fire and seismic risks resulted in total (mean) core damage frequencies of 1.96E-5 and 4.60E-05 per reactor year, respectively. Detailed uncertainty analyses were performed for both fire and seismic risks. These results show that the core damage frequency profile for these events is comparable to that found for existing commercial power plants if proposed fixes are completed as part of the restart program. 108 refs., 85 figs., 80 tabs

Analysis of core damage frequency due to external events at the DOE (Department of Energy) N-Reactor

Energy Technology Data Exchange (ETDEWEB)

Lambright, J.A.; Bohn, M.P.; Daniel, S.L. (Sandia National Labs., Albuquerque, NM (USA)); Baxter, J.T. (Westinghouse Hanford Co., Richland, WA (USA)); Johnson, J.J.; Ravindra, M.K.; Hashimoto, P.O.; Mraz, M.J.; Tong, W.H.; Conoscente, J.P. (EQE, Inc., San Francisco, CA (USA)); Brosseau, D.A. (ERCE, Inc., Albuquerque, NM (USA))

1990-11-01

A complete external events probabilistic risk assessment has been performed for the N-Reactor power plant, making full use of all insights gained during the past ten years' developments in risk assessment methodologies. A detailed screening analysis was performed which showed that all external events had negligible contribution to core damage frequency except fires, seismic events, and external flooding. A limited scope analysis of the external flooding risk indicated that it is not a major risk contributor. Detailed analyses of the fire and seismic risks resulted in total (mean) core damage frequencies of 1.96E-5 and 4.60E-05 per reactor year, respectively. Detailed uncertainty analyses were performed for both fire and seismic risks. These results show that the core damage frequency profile for these events is comparable to that found for existing commercial power plants if proposed fixes are completed as part of the restart program. 108 refs., 85 figs., 80 tabs.

Microbial diversity in a full-scale anaerobic reactor treating high ...

African Journals Online (AJOL)

Microbial characteristics in the up-flow anaerobic sludge blanket reactor (UASB) of a full-scale high concentration cassava alcohol wastewater plant capable of anaerobic hydrocarbon removal were analyzed using cultivation-independent molecular methods. Forty-five bacterial operational taxonomic units (OTUs) and 24 ...

Backflushable filter experience at the N Reactor

International Nuclear Information System (INIS)

Ball, B.; Best, W.T.; Keith, R.C.

1987-01-01

The N Reactor is an 4000 MWt, light-water cooled, graphite-moderated reactor located on the Hanford Site in Washington State. A radwaste pilot plant to process plant effluent was constructed in order to maximize future efficiency when a full size radioactive processing facility is built. The pilot plant's purpose is to vary operational parameters such as filtration and ion exchange on a smaller scale to gather as much data as possible. The input to the pilot plant is radioactive drain lines from the N Reactor. The effluent passes through a backflushable filter and a series of ion exchange columns all scaled down from the future proposed facility. A backflushable filter was selected for this application because of the specific characteristics of the plant effluent and the potential reduced operating costs. The filter performance has been excellent in terms of filtration of the effluent. Typical total suspended solids in the plant effluent range from 1 to 6.1 ppm; the filter reduces this value to less than 0.1 ppm. In addition to outstanding filtration efficiency, the use of a precoat material on the filter has resulted in impressive decontamination factors. The filter has been successful in removing up to 50% of the influent activity. An improved performance of several nuclides over other filtration systems has also been achieved. By varying the composition and amount of precoat material on the filter, substantial reductions in waste volumes (and associated operating and disposal costs) have been demonstrated while maintaining a high degree of removal of both activity and total suspended solids

Feasibility study of full-reactor gas core demonstration test

Science.gov (United States)

Kunze, J. F.; Lofthouse, J. H.; Shaffer, C. J.; Macbeth, P. J.

1973-01-01

Separate studies of nuclear criticality, flow patterns, and thermodynamics for the gas core reactor concept have all given positive indications of its feasibility. However, before serious design for a full scale gas core application can be made, feasibility must be shown for operation with full interaction of the nuclear, thermal, and hydraulic effects. A minimum sized, and hence minimum expense, test arrangement is considered for a full gas core configuration. It is shown that the hydrogen coolant scattering effects dominate the nuclear considerations at elevated temperatures. A cavity diameter of somewhat larger than 4 ft (122 cm) will be needed if temperatures high enough to vaporize uranium are to be achieved.

Substitution of the old console with a fully computerized one. Major maintenance and enhancements of the ventilation and air conditioning system of the reactor building. Research activities at the TRIGA L.E.N.A. plant in Pavia, Italy

International Nuclear Information System (INIS)

Orvini, E.; Piazzoli, A.; Borio, A.

2008-01-01

The TRIGA Mark II reactor of the University of Pavia was operated in the last two years on a routine basis accomplishing different purposes: - Development of B.N.C.T. for diffused tumours of liver; - Neutron activation analysis in matrix of geochemistry, archaeology and environment interest; - Electron spin resonance (ESR) study of radical processes; - Study of trace elements impact in environmental matrix and human health; - High purity control on electronic samples-microchips; - Fast neutron radiation damage investigation; - Certification of standard reference materials and data quality assurance by neutron activation analysis; - Age evaluation by Ar-K method in geological matrix; - Trace elements determination for provenience studies in archaeology; - Basic experiments on fission of Am-242 layers for the space nuclear engine project; - In field search of explosive using prompt gamma emission induced by neutron capture on nitrates. In the period of time between July 1998 and June 2000 the reactor was operated at full power (i.e. 250 kW) for a total amount of 1375 hours. The total fuel element burn-up was 15.07 MWD. During this period of time two major upgrading activities were planned: the installation of a new Instrumentation and Control System (ICS) for the reactor and the installation of a new Air Conditioning and Filtering System (ACFS) for the reactor building. The new ICS is a microprocessor-based design, incorporating the use of one logarithmic wide range neutron flux monitoring channel (NLW-1000), two current mode neutron monitoring safety channels (NP-1000 and NPP-1000) and a linear multi-range neutron flux monitoring channel (Keithley 485). The ICS configuration was personalized by General Atomic according to the requirements and to the technical prescriptions of the Pavia reactor, especially for what concerns the SCRAM inputs and the detectors architecture. Besides, since one the two safety channels, the NPP-1000, has been developed as an advanced

International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2015 edition

International Nuclear Information System (INIS)

Bess, John D.; Gullifor, Jim

2015-03-01

The purpose of the International Reactor Physics Experiment Evaluation (IRPhE) Project is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhE Project is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments', a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The evaluation process entails the following steps: Identify a comprehensive set of reactor physics experimental measurements data, Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, Compile the data into a standardized format, Perform calculations of each experiment with standard reactor physics codes where it would add information, Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at nuclear facilities around the world. The benchmark specifications are intended for use by reactor designers, safety analysts and nuclear data evaluators to validate calculation techniques and data. Example calculations are presented; these do not constitute a validation or endorsement of the codes or cross-section data. The 2015 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments contains data from 143 experimental series that were

Remoción de fósforo de diferentes aguas residuales en reactores aeróbios de lecho fluidizado trifásico con circulación interna

Directory of Open Access Journals (Sweden)

Gleyce Teixeira Correia

2013-01-01

Full Text Available El vertimiento de aguas residuales (AR produce impactos sobre los cuerpos de agua receptores. Nutrientes como P generan implicaciones en los sistemas lénticos pues aceleran los procesos de eutrofización. Se han utilizado diversas tecnologías para la remoción de P de las AR: sistemas de tratamiento físico químico con importantes efectos por adición de productos coagulantes; procesos biológicos basados en alternancia de condiciones anaerobias y aerobias con importantes implicaciones de volumen necesario; sistemas como lagunas de estabilización e irrigación requieren de áreas muy considerables y procesos de postratamiento. Los reactores aerobios de lecho fluidizado con circulación interna (RALFCI son opciones compactas que utilizan gran concentración de biomasa activa que han demostrado su capacidad para remover materia orgánica y N. Para AR domésticas provenientes de la estación de bombeo de Ilha Solteira y para los efluentes de un sistema de recirculación acuícola (SRA de cultivo semi-intensivo de tilapia se evaluó la eficiencia de remoción de P reactivo y P total en tres tipos de RALFCI con diámetro externo de 250 mm y diferentes diámetros de tubo interno (DTI, con dos medios de soporte y diferentes concentraciones en dos de los reactores. Las eficiencias medias de remoción de P reactivo en AR domésticas para un tiempo de retención hidráulica (TRH de 3 horas en el reactor con DTI 125 mm variaron entre 25,6 y 38,4% y en el reactor con DTI 150 mm entre 27,5 y 32,5%; la remoción de P total en el SRA para un TRH de 0,19 h y DTI 100 mm fue de 32,7%.

Safety assessment to support NUE fuel full core implementation in CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Fan, H.Z.; Laurie, T.; Siddiqi, A.; Li, Z.P.; Rouben, D.; Zhu, W.; Lau, V.; Cottrell, C.M. [CANDU Energy Inc., Mississauga, Ontario (Canada)

2013-07-01

The Natural Uranium Equivalent (NUE) fuel contains a combination of recycled uranium and depleted uranium, in such a manner that the resulting mixture is similar to the natural uranium currently used in CANDU® reactors. Based on successful preliminary results of 24 bundles of NUE fuel demonstration irradiation in Qinshan CANDU 6 Unit 1, the NUE full core implementation program has been developed in cooperation with the Third Qinshan Nuclear Power Company and Candu Energy Inc, which has recently received Chinese government policy and funding support from their National-Level Energy Innovation program. This paper presents the safety assessment results to technically support NUE fuel full core implementation in CANDU reactors. (author)

Licensing of MAPLE reactors in Canada

International Nuclear Information System (INIS)

Lee, A.G.; Labrie, J.P.; Langman, V.J.

1999-01-01

Full text: The Operating Licence for a MAPLE reactor (i.e., a 10 MW(th), pool-type reactor), has been approved by the Atomic Energy Control Board (AECB) on August 16th, 1999. This Operating Licence has been obtained within three years of the initiation of the MDS Nordion Medical Isotopes Reactor (MMIR) project, which entails the design, construction and commissioning of two 10 MW MAPLE reactors at AECL's Chalk River Laboratories. The scope and nature of the information required by the AECB, the licensing process and highlights of the events which led to successfully obtaining the Operating Licence for the MAPLE reactor are discussed. These discussions address all phases of the licensing process (i.e., the environmental assessment in support of siting, the Preliminary Safety Analysis Report, PSAR, in support of design, procurement and construction, the Final Safety Analysis Report, FSAR, in support of commissioning and operations, and the development of suitable quality assurance subprograms for each phase). An overview of some of the unique technical aspects associated with the MAPLE reactors, and how they have been addressed during the licensing process are also provided (e.g., applying CSA N285.0, General Requirements for Pressure-Retaining Systems and Components in CANDU Nuclear Power Plants, to a small, low pressure, low temperature research reactor, confirmation of the performance of the driver fuel via laboratory and/or in-reactor testing, validation of the computer codes used to perform the safety analyses, critical parameter uncertainty assessment, full scale hydraulic testing of the performance of the design, fuel handling, human factors validation, operator training and certification). (author)

Development of a system model for advanced small modular reactors.

Energy Technology Data Exchange (ETDEWEB)

Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

2014-01-01

This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

Continuous formation of N-chloro-N,N-dialkylamine solutions in well-mixed meso-scale flow reactors

Science.gov (United States)

Jolley, Katherine E

2015-01-01

Summary The continuous flow synthesis of a range of organic solutions of N,N-dialkyl-N-chloramines is described using either a bespoke meso-scale tubular reactor with static mixers or a continuous stirred tank reactor. Both reactors promote the efficient mixing of a biphasic solution of N,N-dialkylamine in organic solvent, and aqueous sodium hypochlorite to achieve near quantitative conversions, in 72–100% in situ yields, and useful productivities of around 0.05 mol/h with residence times from 3 to 20 minutes. Initial calorimetric studies have been carried out to inform on reaction exotherms, rates and safe operation. Amines which partition mainly in the organic phase require longer reaction times, provided by the CSTR, to compensate for low mass transfer rates in the biphasic system. The green metrics of the reaction have been assessed and compared to existing procedures and have shown the continuous process is improved over previous procedures. The organic solutions of N,N-dialkyl-N-chloramines produced continuously will enable their use in tandem flow reactions with a range of nucleophilic substrates. PMID:26734089

Graphite surveillance in N Reactor

International Nuclear Information System (INIS)

Woodruff, E.M.

1991-09-01

Graphite dimensional changes in N Reactor during its 24 yr operating history are reviewed. Test irradiation results, block measurements, stack profiles, top of reflector motion monitors, and visual observations of distortion are described. 18 refs., 14 figs., 1 tab

Aerobic Sludge Granulation in a Full-Scale Sequencing Batch Reactor

Directory of Open Access Journals (Sweden)

Jun Li

2014-01-01

Full Text Available Aerobic granulation of activated sludge was successfully achieved in a full-scale sequencing batch reactor (SBR with 50,000 m3 d−1 for treating a town’s wastewater. After operation for 337 days, in this full-scale SBR, aerobic granules with an average SVI30 of 47.1 mL g−1, diameter of 0.5 mm, and settling velocity of 42 m h−1 were obtained. Compared to an anaerobic/oxic plug flow (A/O reactor and an oxidation ditch (OD being operated in this wastewater treatment plant, the sludge from full-scale SBR has more compact structure and excellent settling ability. Denaturing gradient gel electrophoresis (DGGE analysis indicated that Flavobacterium sp., uncultured beta proteobacterium, uncultured Aquabacterium sp., and uncultured Leptothrix sp. were just dominant in SBR, whereas uncultured bacteroidetes were only found in A/O and OD. Three kinds of sludge had a high content of protein in extracellular polymeric substances (EPS. X-ray fluorescence (XRF analysis revealed that metal ions and some inorganics from raw wastewater precipitated in sludge acted as core to enhance granulation. Raw wastewater characteristics had a positive effect on the granule formation, but the SBR mode operating with periodic feast-famine, shorter settling time, and no return sludge pump played a crucial role in aerobic sludge granulation.

Epitaxial growth of GaN/AlN/InAlN heterostructures for HEMTs in horizontal MOCVD reactors with different designs

Energy Technology Data Exchange (ETDEWEB)

Tsatsulnikov, A. F., E-mail: andrew@beam.ioffe.ru; Lundin, W. V.; Sakharov, A. V.; Zavarin, E. E.; Usov, S. O.; Nikolaev, A. E.; Yagovkina, M. A.; Ustinov, V. M. [Russian Academy of Sciences, Ioffe Physical–Technical Institute (Russian Federation); Cherkashin, N. A. [CEMES–CNRS—Université de Toulouse (France)

2016-09-15

The epitaxial growth of InAlN layers and GaN/AlN/InAlN heterostructures for HEMTs in growth systems with horizontal reactors of the sizes 1 × 2', 3 × 2', and 6 × 2' is investigated. Studies of the structural properties of the grown InAlN layers and electrophysical parameters of the GaN/AlN/InAlN heterostructures show that the optimal quality of epitaxial growth is attained upon a compromise between the growth conditions for InGaN and AlGaN. A comparison of the epitaxial growth in different reactors shows that optimal conditions are realized in small-scale reactors which make possible the suppression of parasitic reactions in the gas phase. In addition, the size of the reactor should be sufficient to provide highly homogeneous heterostructure parameters over area for the subsequent fabrication of devices. The optimal compositions and thicknesses of the InAlN layer for attaining the highest conductance in GaN/AlN/InAlN transistor heterostructures.

Calculations of radiation defect formation cross sections in reactor materials in (n,p) and (n,α) reactions

International Nuclear Information System (INIS)

Kupchishin, A.A.; Kupchishin, A.I.; Omarbekova, Zh.

2001-01-01

In the work an experimental data analysis by integral σ(E 1 ) and differential [dσ(E 1 ,E 2 )]/dE 2 neutron interaction cross sections with reactor materials with the secondary protons and alpha particles generation as well as with the primarily knock-on atoms production in such reactions are carried out. It is shown, that in the (n,p) and (n',α) reactions the recoil nuclei receive essential energy portion and they are the patriarchs for atom-atom cascades in the substance. Nuclear reactions with formation of the secondary α-particles and and recoil nuclei are considered. It is shown, that these reactions are effectively proceeding within neutrons energy range 0.3-15 MeV. The nuclear reactions kinematics of above mentioned processes is studied. Energy conservation law for these reaction is applied. Deferential cross section conservation and transformation law for radiation defect formation in the (n,α) reaction are considered as well

Measurement of the^ 235U(n,n')^235mU Integral Cross Section in a Pulsed Reactor

Science.gov (United States)

Vieira, D. J.; Bond, E. M.; Belier, G.; Meot, V.; Becker, J. A.; Macri, R. A.; Authier, N.; Hyneck, D.; Jacquet, X.; Jansen, Y.; Legrendre, J.

2009-10-01

We will present the integral measurement of the neutron inelastic cross section of ^235U leading to the 26-minute, E*=76.5 eV isomer state. Small samples (5-20 microgm) of isotope-enriched ^235U were activated in the central cavity of the CALIBAN pulsed reactor at Valduc where a nearly pure fission neutron spectrum is produced with a typical fluence of 3x10^14 n/cm^2. After 30 minutes the samples were removed from the reactor and counted in an electrostatic-deflecting electron spectrometer that was optimized for the detection of ^235mU conversion electrons. From the decay curve analysis of the data, the 26-minute ^235mU component was extracted. Preliminary results will be given and compared to gamma-cascade calculations assuming complete K-mixing or with no K-mixing.

Sedimentación, solubilización, y prefermentación de aguas residuales en un reactor biopelícula

OpenAIRE

Cuevas-Rodríguez, Germán; Tejero Monzón, Iñaki

2003-01-01

Esta investigación fue realizada con el objetivo de desarrollar un nuevo reactor prefermentador de aguas residuales para incrementar los porcentajes de sedimentación, hidrólisis y prefermentación de la materia orgánica contenida en el agua residual bruta, empleando una sola unidad de pretratamiento y, de esta manera, poder remplazar el decantador primario por este nuevo reactor. El estudio fue realizado en un reactor biopelícula de lecho sumergido fijo, empacado con un medio de soporte BLASF‚...

Nuclear energy. The innovations of the N4 reactor

International Nuclear Information System (INIS)

Anon.

1998-01-01

The coupling to the electric network of the two first units of N4 type reactors, on the site of Chooz in the Ardennes, marks the third great step of the French nuclear programme of PWR type reactors, after the realization of 34 units of 900 MWe and 20 units of 1300 M We. The nuclear boiler N4, realizes a new evolution in power, in performances and in reliability. (N.C.)

Efecto de dos metales pesados, cadmio y níquel, sobre la eficiencia de remoción de carga orgánica de un reactor UASB a escala de laboratorio

Directory of Open Access Journals (Sweden)

Luis Eduardo Forero

2004-01-01

Full Text Available Se realizaron ensayos en tres reactores UASB de tres litros cada uno, a un tiempo de retención hidráulico (TRH de cuatro horas y carga orgánica volumétrica de 4,8 g/L/d. Después de la fase inicial de arranque, tiempo de 4.000 horas para los tres reactores, se procedió a afectarlos de la siguiente forma: el primer reactor fue alimentado con 5 mg/L de cloruro de cadmio en forma continua, el segundo reactor fue alimentado con 10 mg/L de cloruro de níquel en forma continua también, mientras que el tercer reactor no se afectó con sustancia alguna y sirvió como control. La eficiencia de remoción de demanda química de oxígeno (DQO del primer reactor cambió del 60% de la fase de arranque (fase 1 al 18% en la fase afectada con cadmio (fase dos; la eficiencia de remoción de DQO en el reactor dos pasó del 60 al 24% y a su vez para el reactor tres control no hubo cambio significativo en dicha eficiencia. A su vez el reactor uno acumuló el cadmio en el lodo, mientras que el reactor dos no hizo lo propio con el níquel.

C-Reactor I and E loading instability limits

Energy Technology Data Exchange (ETDEWEB)

Hess, K.W.

1957-01-24

The pilot charging of I & E fuel elements has been implemented at C-Reactor under Production Test IP-19-A. It was necessary to provide adequate tube protection against flow interruption by establishing proper trip setting on the Panellit pressure gauges. the administration of these Panellit trip settings is done by trip-before- boiling tube outlet temperature limits, which are similar in principle to the current instability limits. Trip-before-boiling limits for C-Reactor I & E fuel elements loadings are presented in this document.

Estudio mecánico e hidrodinámico de un reactor de gasificación de lecho fluidizado.

Directory of Open Access Journals (Sweden)

W. Rosales

2008-01-01

Full Text Available Se realizó una simulación mediante Elementos Finitos y CFD de un prototipo de gasificador experimental a partir de una geometría propuesta. Se abordan aspectos termomecánicos, al calcularse las deformaciones originadas en el equipo, producto de su peso, en las condiciones de emplazamiento y la carga térmica a la que se somete. También se considera el flujo multifásico gas-sólido presente en el lecho fluidizado, se determina el rango de presiones y velocidades de trabajo del dispositivo, y se estudia la evolución del flujo. Para ello se utiliza el modelo de fuerza de arrastre y presión de sólido de Gidaspow, así como los criterios de velocidad mínima de fluidización de Wen & Yu y Kunii & Levenspiel.A Finite Element and CFD simulations were conducted to a prototype of experimental gasifier, starting from a proposed geometry. Thermomechanic aspects are briefed, calculating the reactor deformation, due to its weight and the thermic load. The gas-solid multiphase flow, present on the fluidized bed was also considered, the working range for the pressure and velocity fields were determined and the flow evolution was studied. The drag force and solid pressure models by Gidaspow, and the minimum fluidization velocity criteria, by Wen & Yu and Kunii & Levenspiel were used.

Variability in properties of grouted Phosphate/Sulfate N-Reactor Waste

International Nuclear Information System (INIS)

Lokken, R.O.; Martin, P.F.C.; Bowen, W.M.; Harty, H.; Treat, R.L.

1987-02-01

A Transportable Grout Facility (TGF) is being constructed at the Hanford site in Washington State to convert various low-level liquid wastes to a grout waste form for onsite disposal. The TGF Project is managed by Rockwell Hanford Operations (Rockwell). Oak Ridge National Laboratory (ORNL) has provided a grout formulation for Phosphate/Sulfate N-Reactor Waste, the first waste stream scheduled for grouting beginning in late 1987. The formulation includes a blend of portland cement, fly ash, attapulgite clay, and an illitic clay. Grout will be produced by mixing the blend with Phosphate/Sulfate N-Reactor Waste. These wastes result from decontamination and ion-exchange regeneration activities at Hanford's N-Reactor. Pacific Northwest Laboratory (PNL) is conducting studies on grouted Phosphate/Sulfate N-Reactor Waste to verify that the grout can be successfully processed and, when hardened, that it will meet all performance and regulatory requirements. As part of these studies, PNL is assessing the variability that may be encountered when processing Phosphate/Sulfate N-Reactor Waste grout. Sources of variability that may affect grout properties include the composition and concentrations of the waste and dry solids, temperature, efficiency of dry solids blending, and dry blend storage time. 13 refs., 20 figs., 9 tabs

Fort St. Vrain reactor performance and operation to full power

International Nuclear Information System (INIS)

Simon, W.A.; Bramblett, G.C.

1982-01-01

The Fort St. Vrain Nuclear Generating Station, powered by a high-temperature gas-cooled reactor (HTGR), has now been tested to full thermal power. Testing was conducted for the dual purposes of demonstrating component and system capability as a part of the rise-to-power program and determining core fluctuation/redistribution behavior under full power conditions. Both objectives were met. Full power performance of all major components and the achievement of nearly all design objectives has been verified. In addition, the tests showed that the fluctuation phenomenon has been corrected. Core region outlet temperature redistributions have been characterized, related to a physical mechanism, and shown to be inconsequential for overall plant operation

Methodology used to calculate moderator-system heat load at full power and during reactor transients in CANDU reactors

International Nuclear Information System (INIS)

Aydogdu, K.

1998-01-01

Nine components determine the moderator-system heat load during full-power operation and during a reactor power transient in a CANDU reactor. The components that contribute to the total moderator-system heat load at any time consist of the heat generated in the calandria tubes, guide tubes and reactivity mechanisms, moderator and reflector; the heat transferred from calandria shell, the inner tubesheets and the fuel channels; and the heat gained from moderator pumps and heat lost from piping. The contributions from each of these components will vary with time during a reactor transient. The sources of heat that arise from the deposition of nuclear energy can be divided into two categories, viz., a) the neutronic component (which is directly proportional to neutronic power), which includes neutron energy absorption, prompt-fission gamma absorption and capture gamma absorption; and b) the fission-product decay-gamma component, which also varies with time after initiation of the transient. An equation was derived to calculate transient heat loads to the moderator. The equation includes two independent variables that are the neutronic power and fission-product decay-gamma power fractions during the transient and a constant term that represents the heat gained from moderator pumps and heat lost from piping. The calculated heat load in the moderator during steady-state full-power operation for a CANDU 6 reactor was compared with available measurements from the Point Lepreau, Wolsong 1 and Gentilly-2 nuclear generating stations. The calculated and measured values were in reasonably good agreement. (author)

Proposed high throughput electrorefining treatment for spent N- Reactor fuel

International Nuclear Information System (INIS)

Gay, E.C.; Miller, W.E.; Laidler, J.J.

1996-01-01

A high-throughput electrorefining process is being adapted to treat spent N-Reactor fuel for ultimate disposal in a geologic repository. Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the type of fragmentation necessary to provide fuel segments suitable for this process. Based on these tests, a conceptual design was produced of a plant-scale electrorefiner. In this design, the diameter of an electrode assembly is about 1.07 m (42 in.). Three of these assemblies in an electrorefiner would accommodate a 3-metric-ton batch of N-Reactor fuel that would be processed at a rate of 42 kg of uranium per hour

International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2017 edition

International Nuclear Information System (INIS)

2017-01-01

The International Reactor Physics Evaluation (IRPhE) Project was initiated as a pilot in 1999 by the Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June 2003. While the NEA co-ordinates and administers the IRPhE Project at the international level, each participating country is responsible for the administration, technical direction and priorities of the project within their respective countries. The information and data included in this handbook are available to NEA member countries, to all contributing countries and to others on a case-by-case basis. The IRPhE Project is patterned after the International Criticality Safety Benchmark Evaluation Project (ICSBEP). It closely co-ordinates with the ICSBEP to avoid duplication of efforts and publication of conflicting information. Some benchmark data are applicable to both nuclear criticality safety and reactor physics technology. Some have already been evaluated and published by the ICSBEP, but have been extended to include other types of measurements in addition to the critical configuration. Through this effort, the IRPhE Project will be able to 1) consolidate and preserve the existing worldwide information base; 2) retrieve lost data; 3) identify areas where more data are needed; 4) draw upon the resources of the international reactor physics community to help fill knowledge gaps; 5) identify discrepancies between calculations and experiments due to deficiencies in reported experimental data, cross-section data, cross-section processing codes and neutronics codes; 6) eliminate a large amount of redundant research and processing of reactor physics experiment data, and 7) improve future experimental planning, execution and reporting. This handbook contains reactor physics benchmark specifications that have been derived from experiments performed at nuclear facilities around the world. The benchmark specifications are intended for use by

The Hanford Site N Reactor buildings task identification and evaluation of historic properties

International Nuclear Information System (INIS)

Stapp, D.C.; Marceau, T.E.

1996-01-01

The New Production Reactor complex at Hanford (hereafter referred to as N Reactor) is proposed to be deactivated, decommissioned, and demolished in the coming years. Recognizing that the Hanford Site has been important to the nation, state, and local community, a task was funded to examine the effects that these activities may have on the historic properties of N Reactor. The objectives of the N Reactor buildings task were to identify potential historic properties at N Reactor, to complete Historic Property Inventory forms for all structures considered eligible and ineligible for listing in the National Register of Historic Places, and to prepare a Memorandum of Agreement that identifies the measures required to mitigate any adverse effects

N Reactor Lessons Learned workshop

International Nuclear Information System (INIS)

Heaberlin, S.W.

1993-07-01

This report describes a workshop designed to introduce participants to a process, or model, for adapting LWR Safety Standards and Analysis Methods for use on rector designs significantly different than LWR. The focus of the workshop is on the ''Lessons Learned'' from the multi-year experience in the operation of N Reactor and the efforts to adapt the safety standards developed for commercial light water reactors to a graphite moderated, water cooled, channel type reactor. It must be recognized that the objective of the workshop is to introduce the participants to the operation of a non-LWR in a LWR regulatory world. The total scope of this topic would take weeks to provide a through overview. The objective of this workshop is to provide an introduction and hopefully establish a means to develop a longer term dialogue for technical exchange. This report provides outline of the workshop, a proposed schedule of the workshop, and a description of the tasks will be required to achieve successful completion of the project

MLR reactor

International Nuclear Information System (INIS)

Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.

1998-01-01

The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)

Estudio de distribución de tiempos de residencia en un reactor biológico de lecho empacado cerámico

Directory of Open Access Journals (Sweden)

Tatiana Rodríguez Chaparro

2004-01-01

Full Text Available La distribución de tiempos de residencia de un reactor es una característica del mezclado que ocurre dentro de él [ ] 1 [ ] 2 ; su determinación es básica para el diseño de cualquier tipo de reactor en escala real. El objetivo de esta investigación consistió en determinar la distribución de tiempos de residencia en un reactor biológico de lecho empacado cerámico (anillos a partir de pruebas con trazadores. Los resultados obtenidos utilizando las técnicas de inyección por paso y pulso fueron 34.577 seg., y 17.745 seg., respectivamente, y la dispersión calculada infinita. Lo anterior permite concluir que en reactores de lecho empacado cerámico (anillos las moléculas del trazador se distribuyen uniformemente en todo el sistema. Los ensayos se realizaron en un modelo a escala laboratorio.

E-PROCUREMENT: IMPORTANCIA Y APLICACIÓN

Directory of Open Access Journals (Sweden)

Isabel Fernández Quesada

2003-04-01

Full Text Available En el trabajo se aborda uno de los aspectos externos del e-business: el e-procurement. Se valora la importancia de la utilización de la red en general y en la función logística y de compras, en particular. Partiendo de las actividades tradicionales de compras, se explicitan las ventajas que supone respecto a ellas el e-procurement y la importancia estratégica de su aplicación. Asimismo, se profundiza en el propio concepto de e-procurement y en su arquitectura desde el punto de vista de su implantación empresarial.

N13 - based reactor coolant pressure boundary leakage system

International Nuclear Information System (INIS)

Dissing, E.; Marbaeck, L.; Sandell, S.; Svansson, L.

1980-05-01

A system for the monitoring of leakage of coolant from the reactor coolant pressure boundary and auxiliary systems to the reactor containment, based on the detection of the N13 content in the atmosphere, has been tested. N13 is produced from the oxyegen of the reactor water via the recoil photon nuclear process H1 + 016 + He4. The generation of N13 is therefore independent of fuel element leakage and of the corrosion product content in the water. In the US AEC regulatory guide 1.45 has a leakage increase of 4 liter/ min been suggested as the response limit. The experiments carried out in Ringhals indicate, that with the accomplishment of minor improvements in the installation, a 4 liter/min leakage to the containment will give rise to a signal with a random error range of +- 0.25 liter/min, 99.7 % confidence level. (author)

Microbial Community Composition and Ultrastructure of Granules from a Full-Scale Anammox Reactor

KAUST Repository

Gonzalez-Gil, Graciela

2014-12-11

Granules in anammox reactors contain besides anammox bacteria other microbial communities whose identity and relationship with the anammox bacteria are not well understood. High calcium concentrations are often supplied to anammox reactors to obtain sufficient bacterial aggregation and biomass retention. The aim of this study was to provide the first characterization of bacterial and archaeal communities in anammox granules from a full-scale anammox reactor and to explore on the possible role of calcium in such aggregates. High magnification imaging using backscattered electrons revealed that anammox bacteria may be embedded in calcium phosphate precipitates. Pyrosequencing of 16S rRNA gene fragments showed, besides anammox bacteria (Brocadiacea, 32 %), substantial numbers of heterotrophic bacteria Ignavibacteriacea (18 %) and Anaerolinea (7 %) along with heterotrophic denitrifiers Rhodocyclacea (9 %), Comamonadacea (3 %), and Shewanellacea (3 %) in the granules. It is hypothesized that these bacteria may form a network in which heterotrophic denitrifiers cooperate to achieve a well-functioning denitrification system as they can utilize the nitrate intrinsically produced by the anammox reaction. This network may provide a niche for the proliferation of archaea. Hydrogenotrophic methananogens, which scavenge the key fermentation product H2, were the most abundant archaea detected. Cells resembling the polygon-shaped denitrifying methanotroph Candidatus Methylomirabilis oxyfera were observed by electron microscopy. It is hypothesized that the anammox process in a full-scale reactor triggers various reactions overall leading to efficient denitrification and a sink of carbon as biomass in anammox granules.

Microbial community composition and ultrastructure of granules from a full-scale anammox reactor.

Science.gov (United States)

Gonzalez-Gil, Graciela; Sougrat, Rachid; Behzad, Ali R; Lens, Piet N L; Saikaly, Pascal E

2015-07-01

Granules in anammox reactors contain besides anammox bacteria other microbial communities whose identity and relationship with the anammox bacteria are not well understood. High calcium concentrations are often supplied to anammox reactors to obtain sufficient bacterial aggregation and biomass retention. The aim of this study was to provide the first characterization of bacterial and archaeal communities in anammox granules from a full-scale anammox reactor and to explore on the possible role of calcium in such aggregates. High magnification imaging using backscattered electrons revealed that anammox bacteria may be embedded in calcium phosphate precipitates. Pyrosequencing of 16S rRNA gene fragments showed, besides anammox bacteria (Brocadiacea, 32%), substantial numbers of heterotrophic bacteria Ignavibacteriacea (18%) and Anaerolinea (7%) along with heterotrophic denitrifiers Rhodocyclacea (9%), Comamonadacea (3%), and Shewanellacea (3%) in the granules. It is hypothesized that these bacteria may form a network in which heterotrophic denitrifiers cooperate to achieve a well-functioning denitrification system as they can utilize the nitrate intrinsically produced by the anammox reaction. This network may provide a niche for the proliferation of archaea. Hydrogenotrophic methananogens, which scavenge the key fermentation product H2, were the most abundant archaea detected. Cells resembling the polygon-shaped denitrifying methanotroph Candidatus Methylomirabilis oxyfera were observed by electron microscopy. It is hypothesized that the anammox process in a full-scale reactor triggers various reactions overall leading to efficient denitrification and a sink of carbon as biomass in anammox granules.

N Reactor thermal plume characterization during Pu-only mode of operation

Energy Technology Data Exchange (ETDEWEB)

Ecker, R.M.; Thompson, F.L.; Whelan, G.

1983-04-01

Pacific Northwest Laboratories (PNL) performed field and modeling studies -from March 1982 through June 1983 to characterize the thermal plume from the N Reactor heated water outfall while the N Reactor operated in the Pu-only mode. Part 1 of this report deals with the field studies conducted to characterize the N Reactor thermal plume while in the Pu-only mode of operation. It includes a description of the study area, a description of field tasks and procedures, and data collection results and discussion. Part 2 describes the computer simulation of the thermal plume under different flow conditions and the calibration of the model used. It includes a description of the computer model and the assumptions on which it is based, a presentation of the input data used in this application, and a discussion of modeling results. Because the field studies were restricted by the NPOES permit variance to the spring months when high Columbia River flows prevail the mathematical modeling of the N Reactor thermal plume while the reactor operates in the Pu-only mode is instrumental in characterizing the plume during low Columbia River flows.

Methanogenic population dynamics during startup of a full-scale anaerobic sequencing batch reactor treating swine waste.

Science.gov (United States)

Angenent, Largus T; Sung, Shihwu; Raskin, Lutgarde

2002-11-01

Changes in methanogenic population levels were followed during startup of a full-scale, farm-based anaerobic sequencing batch reactor (ASBR) and these changes were linked to operational and performance data. The ASBR was inoculated with anaerobic digester sludge from a municipal wastewater treatment facility. During an acclimation period of approximately 3 months, the ASBR content was diluted to maintain a total ammonia-N level of approximately 2000mg l(-1). After this acclimation period, the volatile solids loading rate was increased to its design value of 1.7g l(-1) day(-1) with a 15-day hydraulic retention time, which increased the total ammonia-N level in the ASBR to approximately 3,600 mg l(-1). The 16S ribosomal RNA (rRNA) levels of the acetate-utilizing methanogens of the genus Methanosarcina decreased from 3.8% to 1.2% (expressed as a percentage of the total 16S rRNA levels) during this period, while the 16S rRNA levels of Methanosaeta concilii remained low (below 2.2%). Methane production and reactor performance were not affected as the 16S rRNA levels of the hydrogen-utilizing methanogens of the order Methanomicrobiales increased from 2.3% to 7.0%. Hence, it is likely that during operation with high ammonia levels, the major route of methane production is through a syntrophic relationship between acetate-oxidizing bacteria and hydrogen-utilizing methanogens. Anaerobic digestion at total ammonia-N levels exceeding 3500mg l(-1) was sustainable apparently due to the acclimation of hydrogen-utilizing methanogens to high ammonia levels.

Design of first reactor protection system prototype for C A R E M reactor

International Nuclear Information System (INIS)

Azcona, A; Lorenzo, G.; Maciel, F.; Fittipaldi, A

2006-01-01

In this paper we present the design of a prototype of the C A R E M Reactor Protection System, which is implemented on a basis of the digital platform T E L E P E R M X S.The proposed architecture for the Reactor Protection System (R P S) has 4 redundant trains composed by a complete set of sensors, a data acquisition computer and a processing computer.The information from the 4 processing computers goes into to a two voting units with a two out of four (2004) logic and its outputs are combined by a final actuation logic with a voting scheme of one out of two (1002).The prototype is implemented with a unique train.The train inputs are simulated by an Automatic Testing Unit.The pre-established test case or procedure results are fed back into the A T U.The choice of the digital platform T E L E P E R M X S for the R P S implementation allows versatility in the design stage and permits the prototype expansion due to its modular characteristic and the software tools flexibility [es

Searching for full power control rod patterns in a boiling water reactor using genetic algorithms

Energy Technology Data Exchange (ETDEWEB)

Montes, Jose Luis [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jlmt@nuclear.inin.mx; Ortiz, Juan Jose [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jjortiz@nuclear.inin.mx; Requena, Ignacio [Departamento Ciencias Computacion e I.A. ETSII, Informatica, Universidad de Granada, C. Daniel Saucedo Aranda s/n. 18071 Granada (Spain)]. E-mail: requena@decsai.ugr.es; Perusquia, Raul [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: rpc@nuclear.inin.mx

2004-11-01

One of the most important questions related to both safety and economic aspects in a nuclear power reactor operation, is without any doubt its reactivity control. During normal operation of a boiling water reactor, the reactivity control of its core is strongly determined by control rods patterns efficiency. In this paper, GACRP system is proposed based on the concepts of genetic algorithms for full power control rod patterns search. This system was carried out using LVNPP transition cycle characteristics, being applied too to an equilibrium cycle. Several operation scenarios, including core water flow variation throughout the cycle and different target axial power distributions, are considered. Genetic algorithm fitness function includes reactor security parameters, such as MLHGR, MCPR, reactor k{sub eff} and axial power density.

N Reactor pressure tube 1350 postirradiation examination

International Nuclear Information System (INIS)

Cook, D.J.

1977-01-01

The N Reactor pressure tubes were fabricated from Zircaloy-2 primarily due to the excellent corrosion resistance, low neutron absorption, and high strength properties of this alloy. Irradiation damage mechanisms increase the strength and decrease the ductility of the Zircaloy-2. Irradiation data available at the time the tubes were installed indicated that fast neutron irradiation damage mechanisms would not decrease the ductility to unacceptable levels over the estimated plant life of 25 to 30 years. However, because the tubes are a primary coolant system component and only limited data are available on irradiation effects at high fluences, a Postirradiation Examination (PIE) program was developed to assure that service factors do not compromise pressure tube integrity essential to reactor safety. The PIE program requires that a pressure tube be periodically removed from the reactor for destructive testing. The N Reactor Technical Specifications specify that the frequency of pressure tube removal and examination be based upon the previous PIE test results. Four pressure tubes were examined before tube 1350, and the test results were summarized in individual reports. PIE results on tube 1350 were summarized along with the test results on the previous four tubes in a previous report. The purpose of this report is to present in detail the results on PIE of pressure tube 1350, and, in particular, document the technique by which the fracture toughness of the pressure tube was determined

Estudio de un reactor catalítico para la obtención de gas de síntesis

OpenAIRE

Romero Sayago, Sara Isabel

2016-01-01

Este trabajo se centra en el estudio del proceso de reformado de gas natural con vapor de agua para producir gas de síntesis. Un compuesto, que como su nombre indica, es de gran importancia en la síntesis de muchos productos. En concreto, se estudia el reactor heterogéneo catalítico donde tiene lugar la reacción de reformado. Mediante un programa de simulación de procesos químicos, se optimiza el proceso de reformado para obtener un rendimiento elevado en el reactor con el mínimo consumo e...

Auditable Safety Analysis and Final Hazard Classification for the 105-N Reactor Zone and 109-N Steam Generator Zone Facility

International Nuclear Information System (INIS)

Kloster, G.L.

1998-07-01

This document is a graded auditable safety analysis (ASA) and final hazard classification (FHC) for the Reactor/Steam Generator Zone Segment. The Reactor/Steam Generator Zone Segment, part of the N Reactor Complex, that is also known as the Reactor Building and Steam Generator Cells. The installation of the modifications described within to support surveillance and maintenance activities are to be completed by July 1, 1999. The surveillance and maintenance activities addressed within are assumed to continue for the next 15- 20 years, until the initiation of facility D ampersand D (i.e., Interim Safe Storage). The graded ASA in this document is in accordance with EDPI-4.30-01, Rev. 1, Safety Analysis Documentation, (BHI-DE-1) and is consistent with guidance provided by the U.S. Department of Energy. This ASA describes the hazards within the facility and evaluates the adequacy of the measures taken to reduce, control, or mitigate the identified hazards. This document also serves as the FHC for the Reactor/Steam Generator Zone Segment. This FHC is developed through the use of bounding accident analyses that envelope the potential exposures to personnel

Mathematical modeling of nitrous oxide (N2O) emissions from full-scale wastewater treatment plants.

Science.gov (United States)

Ni, Bing-Jie; Ye, Liu; Law, Yingyu; Byers, Craig; Yuan, Zhiguo

2013-07-16

Mathematical modeling of N2O emissions is of great importance toward understanding the whole environmental impact of wastewater treatment systems. However, information on modeling of N2O emissions from full-scale wastewater treatment plants (WWTP) is still sparse. In this work, a mathematical model based on currently known or hypothesized metabolic pathways for N2O productions by heterotrophic denitrifiers and ammonia-oxidizing bacteria (AOB) is developed and calibrated to describe the N2O emissions from full-scale WWTPs. The model described well the dynamic ammonium, nitrite, nitrate, dissolved oxygen (DO) and N2O data collected from both an open oxidation ditch (OD) system with surface aerators and a sequencing batch reactor (SBR) system with bubbling aeration. The obtained kinetic parameters for N2O production are found to be reasonable as the 95% confidence regions of the estimates are all small with mean values approximately at the center. The model is further validated with independent data sets collected from the same two WWTPs. This is the first time that mathematical modeling of N2O emissions is conducted successfully for full-scale WWTPs. While clearly showing that the NH2OH related pathways could well explain N2O production and emission in the two full-scale plants studied, the modeling results do not prove the dominance of the NH2OH pathways in these plants, nor rule out the possibility of AOB denitrification being a potentially dominating pathway in other WWTPs that are designed or operated differently.

Materials surveillance program for C-E NSSS reactor vessels

International Nuclear Information System (INIS)

Koziol, J.J.

1977-01-01

Irradiation surveillance programs for light water NSSS reactor vessels provide the means by which the utility can assess the extent of neutron-induced changes in the reactor vessel materials. These programs are conducted to verify, by direct measurement, the conservatism in the predicted radiation-induced changes and hence the operational parameters (i.e., heat-up, cooldown, and pressurization rates). In addition, such programs provide assurance that the scheduled adjustments in the operational parameters are made with ample margin for safe operation of the plant. During the past 3 years, several documents have been promulgated establishing the criteria for determining both the initial properties of the reactor vessel materials as well as measurement of changes in these initial properties as a result of irradiation. These documents, ASTM E-185-73, ''Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels,'' and Appendix H to 10 CFR 50, ''Reactor Vessel Material Surveillance Program Requirements,'' are complementary to each other. They are the result of a change in the basic philosophy regarding the design and analysis of reactor vessels. In effect, the empirical ''transition temperature approach,'' which was used for design, was replaced by the ''analytical fracture mechanics approach.'' The implementation of this technique was described in Welding Research Council Bulletin 1975 and Appendix G to ASME Code Section III. Further definition of requirements appears in Appendix G to 10 CFR 50 published in July 1973. It is the intent of this paper to describe (1) a typical materials surveillance program for the reactor vessel of a Combustion Engineering NSSS, and (2) how the results of such programs, as well as experimental programs provide feed-back for improvement of materials to enhance their radiation resistance and thereby further improve the safety and reliability of future plants. (author)

Reactores aeróbios de lecho fluidizado trifásico con circulación interna: caracterización hidrodinámica y del soporte

Directory of Open Access Journals (Sweden)

Iván Andrés Sánchez Ortiz

2010-01-01

Full Text Available Los reactores aerobios de lecho fluidizado con circulación interna utilizan biomasa activa adherida a un medio soporte (biopelícula, con la ventaja de retener gran concentración de la misma, utilizando poco espacio físico y pudiendo remover, en un mismo reactor, materia carbonácea y nitrogenada. La circulación del líquido ocurre debido a la diferencia de presión hidrostática producida por la inyección de aire en el tubo de subida. En la condición de medio bifásico fueron estudiadas las velocidades de circulación, fracción volumétrica de aire y coeficiente de transferencia de oxígeno, utilizando cuatro configuraciones de reactores. En medio trifásico se evaluaron las velocidades de circulación, la fracción volumétrica de aire y el coeficiente de transferencia de oxígeno en una de dichas configuraciones. También se realizó la caracterización de tres posibles medios de soporte granulares. Se determinó que la relación de áreas entre los tubos interno y externo y la concentración de medio soporte influencian los valores de velocidad de circulación del líquido, de la fracción volumétrica de aire y de la transferencia de oxígeno y se observaron y compararon características importantes de los medios de soporte.

Application of the new GeN-Foam multi-physics solver to the European sodium fast reactor and verification against available codes - 15226

International Nuclear Information System (INIS)

Fiorina, C.; Mikityuk, K.

2015-01-01

A new multi-physics solver for nuclear reactor analysis, named GeN-Foam (Generalized Nuclear Foam), has been developed by the FAST group at the Paul Scherrer Institut. It is based on OpenFOAM and has been developed for the multi-physics transient analyses of pin-based (e.g., liquid metal Fast Reactors, Light Water Reactors) or homogeneous (e.g., fast spectrum Molten Salt Reactors) nuclear reactors. It includes solutions of coarse or fine mesh thermal-hydraulics, thermal-mechanics and neutron diffusion. In particular, thermal-hydraulics solution can combine on the same mesh both a traditional RANS model and a porous medium model, depending on the desired degree of approximation for each region. In case the active reactor core is modeled as a porous medium, a simple sub-solver computes the sub-scale radial temperature profiles in fuel and cladding. The mesh used for neutronics calculations is deformed according to the displacement field predicted by the thermal-mechanics solver, thus allowing for a direct prediction of expansion-related feedback effects in Fast Reactors. To limit computational requirements, GeN-Foam permits the use of three different unstructured meshes for thermal-hydraulics, thermal-mechanics and neutron diffusion. For the same reason, an adaptive time step is employed. The different equations can be solved altogether or selectively included. In this work, GeN-Foam is applied to the analysis of the European Sodium Fast Reactor (ESFR). In particular, a 3-D model of the ESFR core is set up employing a coarse-mesh porous-medium approach for the thermal-hydraulics. The reactor steady-state and different accidental transients are investigated to offer an overview of GeN-Foam use and capabilities, as well as to preliminarily investigate the impact of a relatively accurate thermal-mechanic treatment on the predicted ESFR behavior. A code-to-code benchmark against the TRACE system code is performed to verify the adequacy of the results provided by the new

Actualización del sistema SCADA y de control para los reactores MQ5 y MQ6 de la planta de Pinturas Condor, Sherwin Williams Ecuador

Directory of Open Access Journals (Sweden)

Jonathan Reinoso

2013-12-01

Full Text Available El presente documento describe la actualización del sistema SCADA para los reactores MQ5 y MQ6 de la planta de Pinturas Condor mediante el software Intouch y la actualización del sistema de control del reactor MQ5 implementado en un controlador lógico programable (PLC de marca SCHNEIDER, además de la arquitectura de control realizada en el proyecto. El sistema SCADA y de control de los reactores permiten la visualización y control de los datos y variables más relevantes durante las diferentes fases de producción de resinas en los reactores MQ5 y MQ6.

International Reactor Physics Experiment Evaluation (IRPhE) Project

International Nuclear Information System (INIS)

2013-01-01

The International Reactor Physics Experiment Evaluation (IRPhE) Project aims to provide the nuclear community with qualified benchmark data sets by collecting reactor physics experimental data from nuclear facilities, worldwide. More specifically the objectives of the expert group are as follows: - maintaining an inventory of the experiments that have been carried out and documented; - archiving the primary documents and data released in computer-readable form; - promoting the use of the format and methods developed and seek to have them adopted as a standard. For those experiments where interest and priority is expressed by member countries or working parties and executive groups within the NEA provide guidance or co-ordination in: - compiling experiments into a standard international agreed format; - verifying the data, to the extent possible, by reviewing original and subsequently revised documentation, and by consulting with the experimenters or individuals who are familiar with the experimenters or the experimental facility; - analysing and interpreting the experiments with current state-of-the-art methods; - publishing electronically the benchmark evaluations. The expert group will: - identify gaps in data and provide guidance on priorities for future experiments; - involve the young generation (Masters and PhD students and young researchers) to find an effective way of transferring know-how in experimental techniques and analysis methods; - provide a tool for improved exploitation of completed experiments for Generation IV reactors; - coordinate closely its work with other NSC experimental work groups in particular the International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Shielding Integral Benchmark Experiment Data Base (SINBAD) and others, e.g. knowledge preservation in fast reactors of the IAEA, the ANS Joint Benchmark Activities; - keep a close link with the working parties on scientific issues of reactor systems (WPRS), the expert

The n-n-bar oscillation experiment at the Pavia reactor

International Nuclear Information System (INIS)

Bressi, G.; Calligarich, E.; Cambiaghi, M.; Dolfini, R.; Gigli, A.; Lanza, A.; Liguori, G.; Piazzoli, A.; Ratti, S.; Torre, P.; Cardarelli, R.; Conversi, M.; De Zorzi, G.; Massa, F.; Santonico, R.; Sebastiani, F.; Zanello, D.; Cesana, A.; Terrani, M.

1986-01-01

The n-n-bar oscillation experiment, presented at the Seventh European Conference of Triga Reactor Users in 1982, has recently entered the data collection phase. The possibility of a neutron-antineutron transition, under particular conditions, is foreseen by some Partial Unified Theory. The aim of the experiment is to detect any transition that should take place or to establish a lower limit for the transition time of at least 10 E7 s. A beam of slow neutrons, after a flight in a pipe, air exhausted and shielded against earth magnetic field, crosses a thin carbon target in which the produced antineutrons annihilate. The annihilation products (charged and neutral pions) are dejected in an apparatus situated all around the target. Due to the relatively low flux available a very large cross section beam (∼ 1 m 2 ) had to be used in order to obtain a neutron intensity adequate to the experimental requirements. This raised several problems concerning radiation protection and shielding, in particular in the detector region. The final experimental set up is described and the results concerning the shielding effectiveness and the intensity of the neutron beam obtained are compared with the values foreseen by computer code calculations. (author)

Determination of the boron content in polyethylene samples using the reactor Orphée

CERN Document Server

Gunsing, F; Aberle, O

2017-01-01

The boron content of two unknown types of polyethylene has been determined relative to a known reference type. Samples of polyethylene, including a known boron-less one, were irradiated with thermal neutrons at the reactor Orphée at Saclay in France. Prompt gamma rays were measured with a CeBr$_3$ detector and the intensity of the 478~keV line from $^{10}${B}(n,$\\alpha_1\\gamma$)$^{7}{Li*} was extracted.

Reactor as furnace and reactor as lamp

International Nuclear Information System (INIS)

Goldanskii, V.I.

1992-01-01

There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

Fish distribution studies near N Reactor, Summer 1983

Energy Technology Data Exchange (ETDEWEB)

Dauble, D.D.; Page, T.L.

1984-06-01

This report summarizes field studies that were initiated in July 1983 to provide estimates of the relative distribution of late-summer outmigrant juvenile salmonids and juvenile resident fish upstream of the N Reactor 009 Outfall. Chinook salmon are among the fish species most sensitive to thermal effects, and impacts to the juvenile outmigrant populations are of particular concern to state and federal regulatory and fisheries management agencies. Therefore, the distribution studies were conducted from late July through September, a period when high ambient river temperatures and low river flows make these salmonid populations most susceptible to thermal effects. In addition, data were not available on the spatial distribution of outmigrant juvenile chinook salmon in late summer. Information on the relative distribution of resident fish populations was also gathered. Previous studies of midstream distribution of juvenile resident fish were limited to a description of ichthyoplankton populations (Beak Consultants, Inc. 1980 Page et al. 1982), and no data were available on vertical or horizontal distribution of juvenile resident fish species near N Reactor. Relative densities and spatial distribution estimates of juvenile salmonid and resident fish species will be used in conjunction with laboratory thermal effects studies (Neitzel et al. 1984) and with plume characterization studies (Ecker et al. 1983) to assess potential impacts of thermal discharge on fish populations near N Reactor.

Review of Past Reactor Activities in Italy, April 1975

International Nuclear Information System (INIS)

Pierantoni, P.

1975-01-01

Since the last meeting of the International Working Group on Fast Reactors the natural vocation of the italian program to a cooperation, illustrated on many occasions, has found factual confirmation. As well-known, a broad partnership with France has been entered into. Four agreements have been concluded, towards May-June 1974 - between C.E.A. and C.N.E.N. for the ulterior development of the French filière and the construction and use of PEC; - between C.E.A. and NIRA for the license on the system; - between C.E.A. and AGIP-NUCLEARE for the fabrication of fuel for fast reactors; - between TECHNICATOME and NIRA for the preparation of the offer and the later construction of Super-Phenix. Those agreements are in the full logic of the previous one passed between the producers E.d.F., ENEL and R.W.E. for the construction of the first two commercial-size plants, extrapolated from Phenix and S.N.R. 300. In particular the NERSA society, acting as client for Super-Phenix, has been founded by those same producers and will receive the offer now prepared by a joint team. The new orientation of the programme and the validity of the previous agreements for its implementation have been confirmed at governmental level by the Interministerial Economic Planning Committee (CIPE), which has also approved the third CNEN's five years Plan

Estudio mecánico e hidrodinámico de un reactor de gasificación de lecho fluidizado. // Mechanic and hydrodynamic study of a fluidized bed gasification reactor.

Directory of Open Access Journals (Sweden)

W. Rosales

2008-01-01

Full Text Available Se realizó una simulación mediante Elementos Finitos y CFD de un prototipo de gasificador experimental a partir de unageometría propuesta. Se abordan aspectos termomecánicos, al calcularse las deformaciones originadas en el equipo,producto de su peso, en las condiciones de emplazamiento y la carga térmica a la que se somete. También se considera elflujo multifásico gas-sólido presente en el lecho fluidizado, se determina el rango de presiones y velocidades de trabajo deldispositivo, y se estudia la evolución del flujo. Para ello se utiliza el modelo de fuerza de arrastre y presión de sólido deGidaspow, así como los criterios de velocidad mínima de fluidización de Wen & Yu y Kunii & Levenspiel.Palabras claves: Gasificación, lecho fluidizado, CFD, FEM, flujo multifásico.______________________________________________________________________________Abstract.A Finite Element and CFD simulations were conducted to a prototype of experimental gasifier, starting from a proposedgeometry. Thermomechanic aspects are briefed, calculating the reactor deformation, due to its weight and the thermic load.The gas-solid multiphase flow, present on the fluidized bed was also considered, the working range for the pressure andvelocity fields were determined and the flow evolution was studied. The drag force and solid pressure models byGidaspow, and the minimum fluidization velocity criteria, by Wen & Yu and Kunii & Levenspiel were used.Key words: gasification, fluidized bed, CFD, Finite Element, Multiphase Flow.

EVOLUCIÓN DEL COMPORTAMIENTO VISCOELÁSTICO DEL ASFALTO INDUCIDA BAJO TERMO-OXIDACIÓN IN SITU EN UN REO-REACTOR

Directory of Open Access Journals (Sweden)

XIOMARA VARGAS

2008-01-01

Full Text Available En este artículo se presentan los resultados del proceso de termo-oxidación de asfalto realizados por primera vez en un reo-reactor. El comportamiento viscoelástico del asfalto pudo ser representado por una ley de potencia (G'(w - wn, G'' (w - w1. La variación del exponente 'n' reflejó los cambios estructurales del asfalto inducidos por el proceso de termo-oxidación. En el intervalo de frecuencia experimental y a 200 y 250°C, los módulos elástico G' y viscoso G'' mostraron una relación del tipo: G'' (w - wn y G' (w ~ wn, este comportamiento es equivalente a un 'gel-débil' y confirma los cambios estructurales del asfalto inducidos por el envejecimiento termo-oxidativo.

COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

OpenAIRE

ALEKSEY. L. IZHUTOV; VALERIY. V. IAKOVLEV; ANDREY. E. NOVOSELOV; VLADIMIR. A. STARKOV; ALEKSEY. A. SHELDYAKOV; VALERIY. YU. SHISHIN; VLADIMIR. M. KOSENKOV; ALEKSANDR. V. VATULIN; IRINA. V. DOBRIKOVA; VLADIMIR. B. SUPRUN; GENNADIY. V. KULAKOV

2013-01-01

The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; th...

INTERPRETACIÓN E INTEGRACIÓN NORMATIVA

Directory of Open Access Journals (Sweden)

Eleonora del Pilar Salazar Londoño

2012-01-01

Full Text Available El objetivo del presente artículo es reflexionar sobre el fenómeno de la interpretación en la norma jurídica. Por esta razón, se construye un diálogo con diversos horizontes teóricos que se sitúan entre los teóricos del derecho de un pensamiento iuspositivista e iusnaturalista, dentro de teorías constructivistas, como Dworkin y Hart, teoría argumentativa como Alexy y Habermas, positivistas puros como Kelsen, entre otros. Todos alimentan el discurso hermenéutico con nuevos conceptos de interpretación, como nuevas tendencias de interpretaciones jurídicas. El diálogo entre los teóricos se direcciona a través de preguntas para resolver el problema de la interpretación en el tema de los principios y la integración normativa.

Environmental assessment for the deactivation of the N Reactor facilities. Revision 1

International Nuclear Information System (INIS)

1994-11-01

This environmental assessment (EA) provides information for the US Department of Energy (DOE) to decide whether the Proposed Action for the N Reactor facilities warrants a Finding of No Significant Impact or requires the preparation of an environmental impact statement (EIS). The EA describes current conditions at the N Reactor facilities, the need to take action at the facilities, the elements of the Proposed Action and alternatives, and the potential environmental impacts. The N Reactor facilities are currently in a surveillance and maintenance program, and will eventually be decontaminated and decommissioned (D and D). Operation and maintenance of the facilities resulted in conditions that could adversely impact human health or the environment if left as is until final D and D. The Proposed Action would deactivate the facilities to remove the conditions that present a potential threat to human health and the environment and to reduce surveillance and maintenance requirements. The action would include surveillance and maintenance after deactivation. Deactivation would take about three years and would involve about 80 facilities. Surveillance and maintenance would continue until final D and D, which is expected to be complete for all facilities except the N Reactor itself by the year 2018

Safety And Transient Analyses For Full Core Conversion Of The Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

2011-01-01

Preparing for full core conversion of Dalat Nuclear Research Reactor (DNRR), safety and transient analyses were carried out to confirm about ability to operate safely of proposed Low Enriched Uranium (LEU) working core. The initial LEU core consisting 92 LEU fuel assemblies and 12 Beryllium rods was analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and fuel cladding fail. Working LEU core response were evaluated under these initial events based on RELAP/Mod3.2 computer code and other supported codes like ORIGEN, MCNP and MACCS2. Obtained results showed that safety of the reactor is maintained for all transients/accidents analyzed. (author)

Overview of JT-60U results toward high integrated performance in reactor-relevant regime

International Nuclear Information System (INIS)

Fujita, T.

2002-01-01

Toward steady sustainment of high integrated performance, we have developed weak magnetic shear (high β p mode) and reversed magnetic shear plasmas. As a large-sized tokamak equipped with a variety of devices for heating, current drive and profile/shape control, JT-60U has high ability to approach the conditions required in reactors: low values of normalized Larmor radius and collisionality, high temperatures with T e > or approx. T i , etc. This paper reviews recent JT-60U results with the emphasis on the projection to the reactor-relevant regime. Full non-inductive current drive has been achieved in a 1.8 MA high β p H-mode plasma with β N 2:4, HH y2 =1.2 and high fusion triple product (3 x 10 20 m -3 keVs) owing to increased N-NB power. In a reversed shear plasma, HH y2 =1.4 at n e /n GW 0.8 under the full non-inductive current drive has been achieved with injection of LHRF and N-NB. In box-type ITBs with reversed shear, barriers for ions and electrons were sustained in a regime with T e > or approx. T i . The pedestal pressure was doubled with increased total poloidal beta in pellet-injected high triangularity plasmas with type I and II ELMs. Stable existence of current hole was demonstrated. (author)

Surface area considerations for corroding N reactor fuel

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Pitner, A.L.

1996-06-01

The N Reactor fuel is corroding at sites where the Zircaloy cladding was damaged when the fuel was discharged from the reactor. Corroding areas are clearly visible on the fuel stored in open cans in the K East Basin. There is a need to estimate the area of the corroding uranium to analyze aspects of fuel behavior as it is transitioned. from current wet storage to dry storage. In this report, the factors that contribute to open-quotes trueclose quotes surface area are analyzed in terms of what is currently known about the N Reactor fuel. Using observations from a visual examinations of the fuel in the K East wet storage facility, a value for the corroding geometric area is estimated. Based on observations of corroding uranium and surface roughness values for other metals, a surface roughness factor is also estimated and applied to the corroding K East fuel to provide an estimated open-quotes trueclose quotes surface area. While the estimated area may be modified as additional data become available from fuel characterization studies, the estimate provides a basis to assess effects of exposed uranium metal surfaces on fuel behavior in operations involved in transitioning from wet to dry storage, during shipment and staging, conditioning, and dry interim storage

Full reactor coolant system chemical decontamination qualification programs

Energy Technology Data Exchange (ETDEWEB)

Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1995-03-01

Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

Energy Technology Data Exchange (ETDEWEB)

Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A. [AREVA - Tour AREVA, 1 Place Jean Millier, 92084 Paris La Defense (France)

2013-07-01

In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

International Nuclear Information System (INIS)

Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A.

2013-01-01

In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO 2 fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory

N reactor external events probabilistic risk assessment

International Nuclear Information System (INIS)

Baxter, J.T.

1989-01-01

An external events probabilistic risk assessment of the N Reactor has been completed. The methods used are those currently being proposed for external events analysis in NUREG-1150. Results are presented for the external hazards that survived preliminary screening. They are earthquake, fire, and external flood. Core damage frequencies for these hazards are shown to be comparable to those for commercial pressurized water reactors. Dominant fire sequences are described and related to 10 CFR 50, Appendix R design requirements. Potential remedial measures that reduce fire core damage risk are described including modifications to fire protection systems, procedure changes, and addition of new administrative controls. Dominant seismic sequences are described. The effect of non-safety support system dependencies on seismic risk is presented

Modelo estadístico para la simulación de reactores de lixiviación ácida

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Mónica Hernández-Rodríguez

2015-05-01

Full Text Available Se desarrolló un modelo estadístico que permite simular el comportamiento de la batería de reactores en el proceso de lixiviación ácida y determinar a partir de parámetros operacionales la eficiencia de extracción de níquel y de cobalto. Al realizar las pruebas de validación se obtuvo que más del 95% de los valores determinados por el modelo están dentro de los límites de confianza estimados, sin embargo se observa una tendencia a que el valor calculado se encuentre por debajo del reportado, lo cual se cumple para el 65,79 % y el 61,84 %, de los datos, con relación a la eficiencia de extracción de níquel y cobalto, respectivamente. Se realizó un análisis de sensibilidad paramétrica para establecer la influencia de las variables de operación en el sistema. Se concluye que la sensibilidad depende del nivel de operación del sistema y que las variables más significativas en todos los niveles son: la concentración de magnesio y la de níquel así como la relación ácido - mineral

Impingement studies at the 100-N reactor water intake

International Nuclear Information System (INIS)

Page, T.L.; Neitzel, D.A.; Gray, R.H.

1977-09-01

Fish impingement and traveling screen passage were studied at the 100-N reactor water intake structure, Columbia River mile 380, from late April to August 1977. Species and numbers of fish affected were determined and compared to those at the adjacent Hanford Generating Project (HGP). Fish protection procedures previously developed for HGP were evaluated for application at 100-N

Development of an attached growth reactor for NH₄-N removal at a drinking water supply system in Kathmandu Valley, Nepal.

Science.gov (United States)

Khanitchaidecha, Wilawan; Shakya, Maneesha; Nakano, Yuichi; Tanaka, Yasuhiro; Kazama, Futaba

2012-01-01

Higher concentrations of ammonium (NH(4)-N) and iron (Fe) than a standard for drinking are typical characteristics of groundwater in the study area. To remove NH(4)-N and Fe, the drinking water supply system in this study consists of a series of treatment units (i.e., aeration and sedimentation, filtration, and chlorination); however, NH(4)-N in treated water is higher than a standard for drinking (i.e., removal efficiency. In accordance with raw groundwater characteristics in the area, effects of low inorganic carbon (IC) and phosphate (PO(4)-P) and high Fe on the removal efficiency were also investigated. The results showed a significant increase in NH(4)-N removal efficiency with reactor length and carrier area. A low IC and PO(4)-P had no effect on NH(4)-N removal, whereas a high Fe decreased the efficiency significantly. The first 550 days operation of a pilot-scale reactor installed in the drinking water supply system showed a gradual increase in the efficiency, reaching to 95-100%, and stability in the performance even with increased flow rate from 210 to 860 L/day. The high efficiency of the present work was indicated because only less than 1 mg of NH(4)-N/L was left over in the treated water.

Synthesis of (E-2,4-Dinitro-N-((2E,4E-4-phenyl-5-(pyrrolidin-1-ylpenta-2,4-dienylideneaniline

Directory of Open Access Journals (Sweden)

Mostafa Fesanghari

2009-07-01

Full Text Available (E-2,4-Dinitro-N-((2E,4E-4-phenyl-5-(pyrrolidin-1-ylpenta-2,4-dienylidene aniline dye was prepared in one pot by reaction of premade N-2,4-dinitrophenyl-3-phenylpyridinium chloride (DNPPC and pyrrolidine in absolute MeOH.

N-reactor charge-discharge system analysis

International Nuclear Information System (INIS)

Tokarz, R.D.; Marr, G.D.; Nesbitt, J.F.

1982-09-01

This report documents an analysis of the existing systems in the N-Reactor fuel flow path. It recommends equipment improvements and changes in that path to allow the charge-discharge rates to be increased to 500 tubes per outage without increasing reactor outage time. The estimated program cost of $14 million is projected over an estimated 3-year period. It does not include costs detailed as part of the existing restoration program or any costs that are considered as normal maintenance. The recommendations contained in this report provide a direction and goal for every critical aspect of the fuel flow path. The way in which these recommendations are implemented may greatly affect the schedule and costs. Previous studies by UNC have shown that enhanced fuel element handling has the potential of increasing productivity by 33 days at a cost benefit estimated at $18 million per year. Enhanced fuel handling provides the greatest potential for productivity improvement of any of the areas considered in these studies

A fuel management study and cycle nuclear design for PW reactors

International Nuclear Information System (INIS)

Minguez, E.; Ahnert, C.; Aragones, J. M.; Corella, M. R.

1975-01-01

A reference reactor was chosen to do a general study involving Fuel Management Evaluations of several cycles, and Design Calculations of cycles already performed, according to a calculation scheme set up in the Reactor Technology Division of the J.E.N., using some computer codes acquired to foreign sources and other ones developed in the J.E.N. (Author) 5 refs

A fuel management study and cycle nuclear design for PW reactors

Energy Technology Data Exchange (ETDEWEB)

Minguez, E; Ahnert, C; Aragones, J M; Corella, M R

1975-07-01

A reference reactor was chosen to do a general study involving Fuel Management Evaluations of several cycles, and Design Calculations of cycles already performed, according to a calculation scheme set up in the Reactor Technology Division of the J.E.N., using some computer codes acquired to foreign sources and other ones developed in the J.E.N. (Author) 5 refs.

Assessment of online monitoring strategies for measuring N2O emissions from full-scale wastewater treatment systems.

Science.gov (United States)

Marques, Ricardo; Rodriguez-Caballero, A; Oehmen, Adrian; Pijuan, Maite

2016-08-01

Clark-Type nitrous oxide (N2O) sensors are routinely used to measure dissolved N2O concentrations in wastewater treatment plants (WWTPs), but have never before been applied to assess gas-phase N2O emissions in full-scale WWTPs. In this study, a full-scale N2O gas sensor was tested and validated for online gas measurements, and assessed with respect to its linearity, temperature dependence, signal saturation and drift prior to full-scale application. The sensor was linear at the concentrations tested (0-422.3, 0-50 and 0-10 ppmv N2O) and had a linear response up to 2750 ppmv N2O. An exponential correlation between temperature and sensor signal was described and predicted using a double exponential equation while the drift did not have a significant influence on the signal. The N2O gas sensor was used for online N2O monitoring in a full-scale sequencing batch reactor (SBR) treating domestic wastewater and results were compared with those obtained by a commercial online gas analyser. Emissions were successfully described by the sensor, being even more accurate than the values given by the commercial analyser at N2O concentrations above 500 ppmv. Data from this gas N2O sensor was also used to validate two models to predict N2O emissions from dissolved N2O measurements, one based on oxygen transfer rate and the other based on superficial velocity of the gas bubble. Using the first model, predictions for N2O emissions agreed by 98.7% with the measured by the gas sensor, while 87.0% similarity was obtained with the second model. This is the first study showing a reliable estimation of gas emissions based on dissolved N2O online data in a full-scale wastewater treatment facility. Copyright © 2016 Elsevier Ltd. All rights reserved.

Commissioning of the Opal reactor cold neutron source

International Nuclear Information System (INIS)

Thiering, R.; Lu, W.; Ullah, R.

2006-01-01

Full text: At OPAL, Australia's first cold neutron facility will form an essential part of the reactor's research programs. Fast neutrons, born in the core of a reactor, interact with a cryogenic material, in this case liquid deuterium, to give them very low energies ( 1 0 m eV). A cold neutron flux of 1.4 1 0 E 1 4 n /cm 2/ s is expected, with a peak in the energy spectrum at 4.2m eV. The cold neutron source reached cryogenic conditions for the first time in late 2005. The cold neutron source operates with a sub-cooled liquid Deuterium moderator at 24 K. The moderator chamber, which contains the deuterium, has been constructed from AlMg 5. The thermosiphon and moderator chamber are cooled by helium gas, in a natural convection thermosiphon loop. The helium refrigeration system utilises the Brayton cycle, and is fully insulated within a high vacuum environment. Despite the proximity of the cold neutron source to the reactor core, it has been considered as effectively separate to the reactor system, due to the design of its special vacuum containment vessel. As OPAL is a multipurpose research reactor, used for beam research as well as radiopharmaceutical production and industrial irradiations, the cold neutron source has been designed with a stand-by mode, to maximise production. The stand-by mode is a warm operating mode using only gaseous deuterium at ambient temperatures (∼ 3 00 K ), allowing for continued reactor operations whilst parts of the cold source are unavailable or in maintenance. This is the first time such a stand-by feature has been incorporated into a cold source facility

Simulation of inlet and outlet riser break sequences in the N Reactor

International Nuclear Information System (INIS)

Bolander, M.A.; Fletcher, C.D.

1988-02-01

This report documents work performed at the Idaho National Engineering Laboratory (INEL) in support of the Westinghouse Hanford Company safety analyses of the N Reactor. The RELAP5/MOD2 computer code was used in analyzing two hypothetical transients. The computer code was modified specifically to simulate the refill behavior in the N Reactor process tubes. The transients analyzed were a double-ended rupture of an inlet riser column and a double-ended rupture of an outlet riser column

Una visión global del e.Learning y su e.Market

Directory of Open Access Journals (Sweden)

Guillermo Vázquez

Full Text Available Las Tecnologías de la Información y de la Comunicación y su aplicación en la formación y entrenamiento de los profesionales, mediante el eLearning, supone una de las grandes revoluciones que caracterizan a la Sociedad del Conocimiento. Sin embargo solo una estrategia global que encare todas sus facetas simultáneamente puede acompasar el desarrollo tecnológico y su implementación efectiva. En este artículo los autores subrayan en primer lugar la importancia del liderazgo en la fase actual de implantación y aceptación del eLearning, y los punto claves sobre los que se debe de sustentar. En segundo lugar explican las características de las acciones formativas online, y su engarce en un Campus Virtual. En tercer lugar se explican las características tecnológicas del eLearning. Finalmente recalcan la necesidad de un eMarket para asegurar que lo altos costes de una nueva metodología, son sostenibles.

Una visión global del e.Learning y su e.Market

Directory of Open Access Journals (Sweden)

Guillermo Vázquez

2006-12-01

Full Text Available Las Tecnologías de la Información y de la Comunicación y su aplicación en la formación y entrenamiento de los profesionales, mediante el eLearning, supone una de las grandes revoluciones que caracterizan a la Sociedad del Conocimiento. Sin embargo solo una estrategia global que encare todas sus facetas simultáneamente puede acompasar el desarrollo tecnológico y su implementación efectiva. En este artículo los autores subrayan en primer lugar la importancia del liderazgo en la fase actual de implantación y aceptación del eLearning, y los punto claves sobre los que se debe de sustentar. En segundo lugar explican las características de las acciones formativas online, y su engarce en un Campus Virtual. En tercer lugar se explican las características tecnológicas del eLearning. Finalmente recalcan la necesidad de un eMarket para asegurar que lo altos costes de una nueva metodología, son sostenibles.

E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

Energy Technology Data Exchange (ETDEWEB)

Van Tichelen, Katrien, E-mail: kvtichel@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Mirelli, Fabio, E-mail: fmirelli@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Greco, Matteo, E-mail: mgreco@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Viviani, Giorgia, E-mail: giorgiaviviani@gmail.com [University of Pisa, Lungarno Pacinotti 43, 56126 Pisa (Italy)

2015-08-15

Highlights: • The E-SCAPE facility is a thermal hydraulic scale model of the MYRRHA fast reactor. • The focus is on mixing and stratification in liquid-metal pool-type reactors. • Forced convection, natural convection and the transition are investigated. • Extensive instrumentation allows validation of computational models. • System thermal hydraulic and CFD models have been used for facility design. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant. The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing. The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE. The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

(E-4-Chloro-N-[(E-2-methyl-3-phenylallylidene]aniline

Directory of Open Access Journals (Sweden)

Jim Simpson

2009-02-01

Full Text Available The title Schiff base compound, C16H14ClN, adopts E configurations with respect to both the C=C and C=N bonds. The dihedral angle between the two aromatic rings is 53.27 (4°, while the plane through the C=C—C=N system is inclined at 9.06 (8° to the benzene ring and 44.92 (5° to the chlorobenzene ring. In the crystal structure, weak C—H...Cl and C—H...N hydrogen bonds stack the molecules down the a axis.

Qsub(N) approximation for slowing-down in fast reactors

International Nuclear Information System (INIS)

Rocca-Volmerange, Brigitte.

1976-05-01

An accurate and simple determination of the neutron energy spectra in fast reactors poses several problems. The slowing-down models (Fermi, Wigner, Goertzel-Greuling...) which are different forms of the approximation with order N=0 may prove inaccurate, in spite of recent improvements. A new method of approximation is presented which turns out to be a method of higher order: the Qsub(N) method. It is characterized by a rapid convergence with respect to the order N, by the use of some global parameters to represent the slowing-down and by the expression of the Boltzmann integral equation in a differential formalism. Numerous test verify that, for the order N=2 or 3, the method gives precision equivalent to that of the multigroup numerical integration for the spectra with greatly reduced calculational effort. Furthermore, since the Qsub(N) expressions are a kind of synthesis method, they allow calculation of the spatial Green's function, or the use of collision probabilities to find the flux. Both possibilities have been introduced into existing reactor codes: EXCALIBUR, TRALOR, RE MINEUR... Some applications to multi-zone media (core, blanket, reflector of Masurca pile and exponential slabs) are presented in the isotropic collision approximation. The case of linearly anisotropic collisions is theoretically resolved [fr

Middle English Preposition Twēn(E

Directory of Open Access Journals (Sweden)

Ciszek-Kiliszewska Ewa

2014-12-01

Full Text Available The present paper focuses on the Middle English preposition twēn(e ‘between, among, in between’. The aim of the study is to review the acknowledged etymology of twēn(e as well as to provide its semantics, dialect distribution, complete textual distribution (record of texts employing twēn(e, and absolute token frequency. Moreover, all texts including the preposition twēn(e are subject to an analysis of the whole variety of prepositions meaning ‘between’ and their token frequency in order to establish the proportions of the use of twēn(e and other discussed prepositions, especially the better established preposition betwēn(e in texts employing twēn(e. The study is based on such extensive electronic databases as the Middle English Dictionary online, the Oxford English Dictionary online and the Corpus of Middle English Prose and Verse as well as on a number of complete Middle English texts. The study of the corpus demonstrates the presence of twēn(e and other prepositions meaning ‘between’ also in texts not listed by the Middle English Dictionary online or the Oxford English Dictionary online under appropriate entries, and thus helps to provide a more complete record of texts and authors utilizing twēn(e and the extent of use of twēn(e as compared to other prepositions meaning ‘between’. Moreover, the study demonstrates that also the other discussed prepositions are often not recorded in particular texts by the MED online or the OED online. In more general terms, the paper points out the need for the use of complete texts for the study of historical prepositions.

Activación del topacio natural irradiado por neutrones en el núcleo del reactor RP-10

OpenAIRE

Gómez, J.; Parreño, Fernando; Lázaro, Gerardo; Vela, Mariano

2003-01-01

Se obtuvieron cristales de topacio activados al ser irradiados con neutrones dentro del núcleo del reactor RP-10. La activación depende del flujo de neutrones, por ello se desarrolló portamuestras (canes de irradiación) para absorber que son los causantes de la activación

Operation experience and maintenance at the TRIGA Mark II L.E.N.A. reactor

International Nuclear Information System (INIS)

Gngoli, F.; Berzero, A.; Lana, F.; Rosti, G.; Meloni, S.

2008-01-01

The TRIGA Mark II reactor of the University of Pavia was operated in the last two years on a routine basis, mostly for neutron activation analysis purposes. Moreover the reactor was completely shutdown in the first six months of this year to allow the dismantling of the NADIR experimental setup. The paper presents: - Reactor operation from July 1990 to June 1992; - Reactor users in the time period January 1990 - December 1991; - Specific activities of some radionuclides in the filling materials; - Specific activity of some radionuclides in thermal column materials. Operations related to dismantling of NADIR experimental facility are described. Finally the new thermal column configuration is presented. Starting from the end inside the reactor tank, a graphite layer (35 cm thick) was positioned, followed by a bismuth layer (10 cm thick) to reduce gamma-ray intensity. The old graphite rods were then positioned leaving in the central part, on the equatorial plane of the thermal column, a cavity whose vertical section has 40 cm width and 20 cm height. The bottom of the cavity, towards to the reactor tank, has been lined with additional layers of graphite (10 cm), bismuth (10 cm) and again graphite (1 cm). The new configuration allowed new experiments to be performed. The cavity in the central part has been created to allow the irradiation of large biological samples such as experimental animal and human livers. This is a peculiar step in a neutron capture boron therapy project to be carried out at the University of Pavia. In order to avoid an implemented 41 Ar production in the void space between shutters and the thermal column outer end, the external surface of the thermal column has been coated with boral sheets. The neutron flux profile, both thermal and epithermal, and cadmium ratio for gold are shown. The flux distribution appears to be adequate to proceed with the neutron capture boron therapy experiment. The LENA Health Physics Service has checked all phases of

Comportamiento del acero de baja aleación SA-508 y del acero al carbono A-410b en las condiciones de operación y parada del circuito primario de los reactores de agua ligera tipo PWR

Directory of Open Access Journals (Sweden)

García-Redondo, María del Sol

2000-04-01

Full Text Available The corrosion rate of low alloy steel SA-508 and carbon steel A-410b in simulated operation and shutdown conditions of pressurized water reactor has been determined. Moreover potentiodynamic polarization curves and galvanic effect through coupling of AISI-304 have been carried out under shutdown simulated condition.

En este trabajo se ha determinado la cinética de corrosión del acero de baja aleación SA-508 y del acero al carbono A-410b en condiciones que simulan la operación y la parada de los reactores de agua ligera a presión. También se han realizado curvas de polarización potenciodinámica y se ha estudiado el acoplamiento galvánico con AISI-304 en condiciones de parada de los reactores de agua ligera a presión.

RESOLUCIÓN DE UN PROBLEMA INVERSO VÍA SUBESTIMACIÓN LINEAL A TROZOS //\tAN INVERSE PROBLEM RESOLUTION VIA PIECEWISE LINEAR UNDERESTIMATION

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Javier Hernádez Benítez

2012-12-01

Full Text Available In reactor design phase bubble column type (CBT is required to have the distribution of solids within the reactor. This distribution satisfies an ordinary differential equation (ODE of order two, with boundary conditions that was developed by D. R. Cova [2], followed by D. N. Smith and J. A. Ruether [8]. Some elements of this equation are given by correlations that depend on certain parameters that are unknown but may be obtained from experimental data. The methodology used to determine these parameters is the sub- piecewise linear underestimation developed by O. L. Mangasarian, J. B. Rosen, M. E. Thompson. // RESUMEN: En el diseño de reactores trifásicos tipo columna de burbujeo (CBT, se requiere tener la distribución de solidos dentro del reactor. Esta distribución satisface una ecuación diferencial ordinaria (EDO de orden dos, con condiciones de frontera que fue desarrollada por D. R. Cova [2], y posteriormente por D. N. Smith y J. A. Ruether [8]. Algunos elementos de esta ecuación están dados por correlaciones que dependen de ciertos parámetros que son desconocidos, pero se pueden obtener a partir de datos experimentales. La metodología utilizada para determinar dichos parámetros es la sub-estimación lineal a trozos desarrollada por O. L. Mangasarian, J. B. Rosen y M. E. Thompson.

Development of a pressurizer level compensator for use on N Reactor

International Nuclear Information System (INIS)

Bussell, J.H.

1985-07-01

The instrument described in this report has been developed to compensate the measured water level in the N Reactor pressurizer for temperature effects. N Reactor is a pressurized water nuclear reactor (PWR). The instrument is defined as a pressurizer level compensator (PLC). A pressurizer is used in a PWR to control the primary coolant pressure and provide a surge volume for primary coolant expansion and contraction. A means of compensating for water and steam density is required because of the wide range of pressure and temperature that result from different steady state and transient reactor power levels. The uncompensated level is determined by measurement of differential pressure between the top of the level measurement zone and the bottom of the level measurement zone. Temperature of the water in the pressurizer is the parameter that is used to determine the proper level compensation since water and steam density are primarily functions of temperature in this case. The PLC uses a microprocessor to calculate the compensated level from temperature and differential pressure measurements. This report includes a description of the design, development, and implementation of software and hardware that are in the PLC. 9 refs., 51 figs., 17 tabs

RELACIÓN DE LA INTENCIÓN E IDEACIÓN SUICIDA CON ALGUNAS VARIABLES

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Paola Valderrama

2006-06-01

Full Text Available El objetivo de este estudio fue hallar la relación de la intención e ideación suicida con algunas variables sociodemográficas, las características de la enfermedad, el tratamiento antirretroviral y los aspectos piscoafectivos en personas con el VIH/SIDA que residen en Bogotá (Colombia y pertenecientes a una fundación. Para ello, se tomó una muestra de 75 pacientes con VIH positivo a quienes se evaluó a través de una entrevista semiestructurada y dos instrumentos: el inventario de depresión de Beck (IBD y la escala de Ideación Suicida (SSI. Los resultados mostraron que solo 29 personas tenían ideación suicida y 10 intención suicida: De los 10 solo 2 sujetos tenían intención después del diagnóstico. Se encontró una relación estadística significativa para ideación suicida e intención suicida con edad, estrato socioeconómico, estado civil, orientación sexual, las características de la enfermedad, el tratamiento antirretroviral y los aspectos piscoafectivos.

Technical safety appraisal of the N-Reactor

International Nuclear Information System (INIS)

1986-07-01

This report presents the results of a Technical Safety Appraisal conducted at the Hanford N-Reactor. A team of specialists gathered information for about three weeks on all areas related to safety at the plant. Operational practices, maintenance practices, training drills, and hardware condition were observed. Several recommendations are made in order to correct incomplete rule implementation, to correct hazardous practices, and to promote improvement in satisfactory areas

Environmental assessment for the deactivation of the N Reactor facilities

International Nuclear Information System (INIS)

1995-05-01

This environmental assessment (EA) provides information for the US Department of Energy (DOE) to decide whether the Proposed Action for the N Reactor facilities warrants a Finding of No Significant Impact or requires the preparation of an environmental impact statement (EIS). The EA describes current conditions at the N Reactor facilities, the need to take action at the facilities, the elements of the Proposed Action and alternatives, and the potential environmental impacts. As required by the National Environmental Policy Act of 1969 (NEPA), this EA complies with Title 40, Code of Federal Regulations (CFR), parts 1500--1508, ''Regulations for Implementing the Procedural Provisions of NEPA. '' It also implements the ''National Environmental Policy Act; Implementing Procedures and Guidelines'' (10 CFR 1021)

Reseña. Ética,educación e investigación

Directory of Open Access Journals (Sweden)

Julio César Arboleda

2018-02-01

Full Text Available Algunos de los artículos de este número orbitan alrededor del tema ética, Educación e Investigación. En particular aquellos que reivindican la visión ética y antropológica de la educación, que sigue constituyendo un pasivo en los procesos de reflexión y formación. Es el caso del trabajo editorial a cargo del pedagogo español Pedro Ortega, creador de la Pedagogía de la alteridad y miembro del Comité de calidad de Redipe. Con el profesor Eduardo avanzar una conversación relevante sobre el tema. Los demás artículos se articulan a tópicos de ética e investigación, la actitud, la inclusión y la convivencia, entre otros.

Steady-state thermal-hydraulic analysis of the Moroccan TRIGA MARK II reactor by using PARET/ANL and COOLOD-N2 codes

International Nuclear Information System (INIS)

Boulaich, Y.; Nacir, B.; El Bardouni, T.; Zoubair, M.; El Bakkari, B.; Merroun, O.; El Younoussi, C.; Htet, A.; Boukhal, H.; Chakir, E.

2011-01-01

Research highlights: → The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. → The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). → The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. In order to validate our PARET/ANL and COOLOD-N2 models, the fuel center temperature as function of core power was calculated and compared with the corresponding experimental values. The comparison indicates that the calculated values are in satisfactory agreement with the measurement. The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). Therefore, we have calculated the departure from nucleate boiling ratio (DNBR), fuel center and surface temperature, cladding surface temperature and coolant temperature profiles across the hottest channel. The most important conclusion is that all obtained values are largely far to compromise safety of the reactor.

Color coherent effects in (e,e'N) and (e,e'N,N(h)) processes at CEBAF

International Nuclear Information System (INIS)

Frankfurt, L.L.; Sargsyan, M.M.; Strikman, M.I.

1994-01-01

The options for investigating color coherent effects and competing nuclear effects of nucleon-nucleon correlations in nuclei, nuclear shell effects in (e, e'N) and (e, e'NN(h)) reactions are considered. They argue that extension of CEBAF energies to reach Q 2 = 10 GeV 2 will allow systematical investigations of color coherent effects in nonperturbative regime of QCD and their interplay with nuclear effects

Waste conversion into n-caprylate and n-caproate: resource recovery from wine lees using anaerobic reactor microbiomes and in-line extraction

Directory of Open Access Journals (Sweden)

Leo A. Kucek

2016-11-01

Full Text Available To convert wastes into sustainable liquid fuels and chemicals, new resource recovery technologies are required. Chain elongation is a carboxylate-platform bioprocess that converts short-chain carboxylates (SCCs (e.g., acetate C2 and n-butyrate C4 into medium-chain carboxylates (MCCs (e.g., n-caprylate C8 and n-caproate C6 with hydrogen gas as a side product. Ethanol or another electron donor (e.g., lactate, carbohydrate is required. Competitive MCC productivities, yields (product vs. substrate fed, and specificities (product vs. all products were only achieved previously from an organic waste material when exogenous ethanol had been added. Here, we converted a real organic waste, which inherently comprised of ethanol, into MCCs with n-caprylate as the target product. We used wine lees, which consisted primarily of settled yeast cells and ethanol from wine fermentation, and produced MCCs with a reactor microbiome. We operated the bioreactor at a pH of 5.2 and with continuous in-line extraction and achieved a MCC productivity of 3.9 g COD/L-d at an organic loading rate of 5.8 g COD/L-d, resulting in a promising MCC yield of 67% and specificities of 36% for each n-caprylate and n-caproate (72% for both. Compared to all other studies that used complex organic substrates, we achieved the highest n-caprylate-to-n-caproate product ratio of 1.0 (COD basis, because we used increased broth-recycle rates through the forward membrane contactor, which improved in-line extraction rates. Increased recycle rates also allowed us to achieve the highest reported MCC production flux per membrane surface area thus far (20.1 g COD/m2-d. Through microbial community analyses, we determined that an operational taxonomic unit (OTU for Bacteroides spp. was dominant and was positively correlated with increased MCC productivities. Our data also suggested that the microbiome may have been shaped for improved MCC production by the high broth-recycle rates. Comparable abiotic

Team training using full-scale reactor coolant pump seal mock-ups

International Nuclear Information System (INIS)

McDonald, T.J.; Hamill, R.W.

1987-01-01

The use of full-scale reactor coolant pump (RCP) seal mock-ups has greatly enhanced Northeast Utilities' ability to effectively utilize the team training approach to technical training. With the advent of the Institute of Nuclear Power Operations accreditation come a new emphasis and standards for the integrated training of plant engineering personnel, maintenance mechanics, quality control personnel, and health physics personnel. The results of purchasing full-scale RCP mock-ups to pilot the concept of team training have far exceeded expectations and cost-limiting factors. The initial training program analysis identified RCP seal maintenance as a task that required training for maintenance department personnel. Due to radiation exposure considerations and the unavailability of actual plant equipment for training purposes, the decision was made to procure a mock-up of an RCP seal assembly and housing. This mock-up was designed to facilitate seal cartridge removal, disassembly, assembly, and installation, duplicating all internal components of the seal cartridge and housing area in exact detail

Biomedicalización e infancia: trastorno de déficit de atención e hiperactividad

Directory of Open Access Journals (Sweden)

Celia Iriart

2012-12-01

Full Text Available El artículo analiza críticamente el aumento de los niños diagnosticados y tratados por el Trastorno de Déficit de Atención e Hiperactividad (TDAH. Los análisis vinculan este creciente fenómeno con las estrategias de la industria farmacéutica para reposicionarse en el liderazgo de la conceptualización del proceso salud-enfermedad-atención y en el mercado de salud. Utilizamos métodos analítico-interpretativos para estudiar datos primarios y secundarios, y realizar una extensa revisión bibliográfica. A la luz del concepto de biomedicalización analizamos los mecanismos subjetivo-ideológicos que facilitaron que este discurso se instituya como una nueva verdad sobre este trastorno y sea legitimado por los organismos gubernamentales y las organizaciones de la sociedad civil. La biomedicalización del sufrimiento infantil dificulta que se pongan en evidencia los profundos cambios socioeconómicos, políticos e ideológico-culturales que han transformado radicalmente nuestras sociedades en las últimas décadas.

Analysis of the interim safe storage of reactors at the Hanford site

International Nuclear Information System (INIS)

Wang Hailiang

2014-01-01

The nine production reactors, i.e. B, C, D, DR, F, H, KE, KW and N, at the Hanford site are all water-cooled and graphite-moderated reactors with natural uranium fuel. In 1993, the U.S. Department of Energy (DOE) decided to put eight production reactors (except for B) into Interim Safe Storage (ISS) for 75 years followed by deferred one-piece removal. Reactor B will remain as a national historical landmark. By the end of 2013, six reactors C, F, D, DR, H and N had been successfully put into the ISS. Reactors KE and KW will be put into the ISS in the coming years. Taking reactor C as an example, this paper mainly talks about how to put the production reactors in the Interim Safe Storage, e.g. how to make site preparation, how to construct the safe storage enclosure (SSE) and how to perform surveillance and maintenance during the ISS period, etc. (authors)

Study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP

International Nuclear Information System (INIS)

Soares, Alexandre de Souza

2014-01-01

The present paper proposes a study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP. To this end, it considered a loss of coolant accident with 160 cm 2 breaking area in cold leg of 20 circuit of the reactor cooling system of nuclear power plant Angra 2, with the reactor operating in stationary condition, to 100% power. It considered that occurred at the same time the loss of External Power Supply and the availability of emergency cooling system was not full. The results obtained are quite relevant and with the possibility of being used in the planning of future activities, given that the construction of Angra 3 is underway and resembles the Angra 2. (author)

Effects of C/N ratio on nitrous oxide production from nitrification in a laboratory-scale biological aerated filter reactor.

Science.gov (United States)

He, Qiang; Zhu, Yinying; Fan, Leilei; Ai, Hainan; Huangfu, Xiaoliu; Chen, Mei

2017-03-01

Emission of nitrous oxide (N 2 O) during biological wastewater treatment is of growing concern. This paper reports findings of the effects of carbon/nitrogen (C/N) ratio on N 2 O production rates in a laboratory-scale biological aerated filter (BAF) reactor, focusing on the biofilm during nitrification. Polymerase chain reaction-denaturing gradient gel electrophoresis (PCR-DGGE) and microelectrode technology were utilized to evaluate the mechanisms associated with N 2 O production during wastewater treatment using BAF. Results indicated that the ability of N 2 O emission in biofilm at C/N ratio of 2 was much stronger than at C/N ratios of 5 and 8. PCR-DGGE analysis showed that the microbial community structures differed completely after the acclimatization at tested C/N ratios (i.e., 2, 5, and 8). Measurements of critical parameters including dissolved oxygen, oxidation reduction potential, NH 4 + -N, NO 3 - -N, and NO 2 - -N also demonstrated that the internal micro-environment of the biofilm benefit N 2 O production. DNA analysis showed that Proteobacteria comprised the majority of the bacteria, which might mainly result in N 2 O emission. Based on these results, C/N ratio is one of the parameters that play an important role in the N 2 O emission from the BAF reactors during nitrification.

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-03-01

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

Energy-averaged neutron cross sections of fast-reactor structural materials

International Nuclear Information System (INIS)

Smith, A.; McKnight, R.; Smith, D.

1978-02-01

The status of energy-averaged cross sections of fast-reactor structural materials is outlined with emphasis on U.S. data programs in the neutron-energy range 1-10 MeV. Areas of outstanding accomplishment and significant uncertainty are noted with recommendations for future efforts. Attention is primarily given to the main constituents of stainless steel (e.g., Fe, Ni, and Cr) and, secondarily, to alternate structural materials (e.g., V, Ti, Nb, Mo, Zr). Generally, the mass regions of interest are A approximately 50 to 60 and A approximately 90 to 100. Neutron total and elastic-scattering cross sections are discussed with the implication on the non-elastic-cross sections. Cross sections governing discrete-inelastic-neutron-energy transfers are examined in detail. Cross sections for the reactions (n;p), (n;n',p), (n;α), (n;n',α) and (n;2n') are reviewed in the context of fast-reactor performance and/or diagnostics. The primary orientation of the discussion is experimental with some additional attention to the applications of theory, the problems of evaluation and the data sensitivity of representative fast-reactor systems

EVALUACIÓN DE LA SENSIBILIDAD PARAMÉTRICA DEL PROCESO DE SÍNTESIS DE LA CICLONITA EN UN REACTOR POR LOTES

Directory of Open Access Journals (Sweden)

Juan Carlos Ojeda Toro

Full Text Available En este trabajo se evaluó la sensibilidad paramétrica del proceso de síntesis de la ciclonita en un reactor por lotes. Esto con el fin de definir condiciones seguras de operación para esta reacción altamente exotérmica. La ley de velocidad de la reacción se ajustó a partir de datos experimentales disponibles en la literatura. Reparametrizando los balances de materia y energía del reactor, se estableció la sensibilidad de la temperatura de reacción con respecto a la variación de la temperatura inicial del sistema reactivo y la temperatura del medio refrigerante. Para determinar las condiciones críticas de operación del reactor, se usó como criterio el cálculo de los coeficientes de sensibilidad y los perfiles de temperatura-conversión así como el lugar geométrico de los máximos de estas curvas. Se definió para el sistema reactivo un potencial crítico de generación de calor (M igual a 34 y que las condiciones críticas de Runaway corresponden a un número de Semenov (ψ igual a 0,684, un parámetro de calor de reacción (B igual a 15 y un número del tipo Arrhenius (γ con un valor de 20. Así mismo, los perfiles de temperatura-conversión precisan una relación crítica entre el potencial de enfriamiento y generación de calor de 2,5786 (N/M.

Chargino and neutralino production at e{sup +}e{sup -} colliders in the complex MSSM. A full one-loop analysis

Energy Technology Data Exchange (ETDEWEB)

Heinemeyer, S. [Campus of International Excellence UAM+CSIC, Madrid (Spain); Universidad Autonoma de Madrid, Cantoblanco, Instituto de Fisica Teorica (UAM/CSIC), Madrid (Spain); Instituto de Fisica de Cantabria (CSIC-UC), Santander (Spain); Schappacher, C.

2017-09-15

For the search for charginos and neutralinos in the Minimal Supersymmetric Standard Model (MSSM) as well as for future precision analyses of these particles an accurate knowledge of their production and decay properties is mandatory. We evaluate the cross sections for the chargino and neutralino production at e{sup +}e{sup -} colliders in the MSSM with complex parameters (cMSSM). The evaluation is based on a full one-loop calculation of the production mechanisms e{sup +}e{sup -} → χ{sub c}{sup ±}χ{sub c}{sup {sub '}-+} and e{sup +}e{sup -} → χ{sub n}{sup 0}χ{sub n}{sup {sub '}0} including soft and hard photon radiation. We mostly restricted ourselves to a version of our renormalization scheme which is valid for vertical stroke M{sub 1} vertical stroke < vertical stroke M{sub 2} vertical stroke, vertical stroke μ vertical stroke and M{sub 2} ≠ μ to simplify the analysis, even though we are able to switch to other parameter regions and correspondingly different renormalization schemes. The dependence of the chargino/neutralino cross sections on the relevant cMSSM parameters is analyzed numerically. We find sizable contributions to many production cross sections. They amount to roughly ±15% of the tree-level results but can go up to ±40% or higher in extreme cases. Also the complex phase dependence of the one-loop corrections was found non-negligible. The full one-loop contributions are thus crucial for physics analyses at a future linear e{sup +}e{sup -} collider such as the ILC or CLIC. (orig.)

The Jules Horowitz Reactor project, a driver for revival of the research reactor community

International Nuclear Information System (INIS)

Pere, P.; Cavailler, C.; Pascal, C.

2010-01-01

The first concrete of the nuclear island for the Jules Horowitz Reactor (JHR) was poured at the end of July 2009 and construction is ongoing. The JHR is the largest new platform for irradiation experiments supporting Generation II and III reactors, Generation IV technologies, and radioisotope production. This facility, composed of a unique grouping of workshops, hot cells and hot laboratories together with a first-rate MTR research reactor, will ensure that the process, from preparations for irradiation experiments through post-irradiation non-destructive examination, is completed expediently, efficiently and, of course, safely. In addition to the performance requirements to be met in terms of neutron fluxes on the samples (5x10 14 n.cm -2 /sec -1 E>1 MeV in core and 3,6x10 14 n.cm -2 /sec -1 E<0.625 eV in the reflector) and the JHR's considerable irradiation capabilities (more than 20 experiments and one-tenth of irradiation area for simultaneous radioisotope production), the JHR is the first MTR to be built since the end of the 1960s, making this an especially challenging project. The presentation will provide an overview of the reactor, hot cells and laboratories and an outline of the key milestones in the project schedule, including initial criticality in early 2014 and radioisotope production in 2015. This will be followed by a description of the project organization set up by the CEA as owner and future operator and AREVA TA as prime contractor and supplier of critical systems, and a discussion of project challenges, especially those dealing with the following items:accommodation of a broad experimental domain; involvement by international partners making in-kind contributions to the project; ? development of components critical to safety and performance; the revival of engineering of research reactors and experimental devices involving France's historical players in the field of research reactors, and; tools to carry out the project, including computer codes

Simulación CFD de la transferencia de calor en un reactor de hidrotratamiento de aceites vegetales de segunda generación

OpenAIRE

Mendoza Sépulveda, César Camilo

2013-01-01

Resumen: Se desarrolló un modelo CFD que permite representar la transferencia de calor en un reactor de hidrotratamiento de aceites vegetales. Este modelo permitió evaluar la transferencia de calor para distintas configuraciones del reactor. En el proceso de hidrotratamiento de aceites vegetales se transforma el aceite en un líquido con cero contenido de azufre y excelentes propiedades como combustible diesel. El proceso se basa en la adición de hidrógeno a alta presión en un reactor de lecho...

Análisis de estabilidad del reactor PFTR para una reacción con cinética de primer orden utilizando la funcional de Lyapunov

Directory of Open Access Journals (Sweden)

Héctor Armando Durán Peralta

2007-01-01

Full Text Available Abunda la literatura referente al análisis de estabilidad de reactores con parámetros globalizados de concentración y temperatura (por ejemplo el CSTR, en cambio es escasa la literatura sobre la estabilidad de reactores con parámetros distribuidos donde existe distribución espacial de concentración y temperatura, como es el caso del reactor tubular PFTR. Este documento analiza la estabilidad del reactor PFTR isotérmico y no isotérmico para una reacción con cinética de primer orden utilizando la funcional de Lyapunov. Se trabaja con una cinética de primer orden pues un objetivo de este artículo es mostrar cómo se aplica la funcional de Lyapunov al análisis de un reactor de parámetros distribuidos, dado que es casi inexistente la literatura sobre el método de la funcional de Lyapunov aplicada a la estabilidad de reactores (técnica usada en el análisis de estabilidad de sistemas en ingeniería eléctrica. El análisis de estabilidad dio como resultado perfiles de temperatura y concentración asintóticamente estables para los casos PFTR isotérmico, no isotérmico con constante cinética independiente de la temperatura y PFTR no isotérmico adiabático. Para el PFTR con retiro de calor el análisis condujo a una región de estabilidad asintótica y a una región incierta donde puede o no haber oscilaciones.

N Reactor updated safety analysis report, NUSAR

International Nuclear Information System (INIS)

1978-01-01

An update of the N Reactor safety analysis is presented to reconfirm that the continued operation does not pose undue risk to DOE personnel and property, the public, or the environment. A reanalysis of LOCA and reactivity transients utilizing current codes and methods is made. The principal aspects of the overall submission, a general description, and site characteristics including geography and demography, nearby industrial, transportation and military facilities, meteorology, hydraulic engineering, and geology and seismology are described

Determination of the neutron spectrum at different locations in the Argentine RA-1 Reactor; Determinacion del espectro neutronico en distintas posiciones del reactor RA-1

Energy Technology Data Exchange (ETDEWEB)

Lerner, A M; Madariaga, M R [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

1999-12-31

Full text: It is well known that the RA-1 reactor is used to irradiate different types of materials with neutrons. The Radio dosimetry Group (which belongs to the Nuclear Regulatory Authority) uses its fast column for the design, calibration and set up of criticality dosimeters as well as for a quick assessment of the dose to workers in case of an accident. With such purpose, Au(1), Au under Cd and In(2) foils were irradiated to estimate absolute thermal, epithermal and fast neutron fluxes at the irradiation location. The accuracy of this estimation is higher when the response to the present neutron spectrum of the different materials constituting the detectors is better known. This, in turn, requires the previous knowledge of such spectrum (a detailed energy dependence of neutron flux) at the analysed location. In this work a neutronic calculation is presented at the fast irradiation location. The whole calculation was carried out following two different methodologies, and considering a power of 40 kW. The reactor and its surroundings were represented by a simplified one-dimensional model, as a concentric cylindrical set of regions. Figures are drawn representing fast and thermal fluxes (with the cut at 0.4 eV) as a function of the distance to the core centre. The neutron flux (in n/cm{sup 2}sec.eV) as a function of energy is also shown at the fast irradiation location. Values of flux (in n/cm{sup 2}.sec.eV) are also provided as a function of energy in other typical locations, as well as the equivalent integrated flux values (in n/cm{sup 2}.sec). ((1) According to the reaction Au{sup 197}(n,{gamma})Au{sup 198}, having a cross section of {sigma}{sub 0}=98.8b for thermal neutrons. (2) According to the reaction In{sup 115}(n,n`)In{sup 115m}, with a cross section of some 70 mb for neutrons with energies above 1.2MeV). (author) [Espanol] Texto completo: Como se sabe, el reactor RA1 se utiliza para irradiar con neutrones distintos tipos de materiales. El grupo de

Uso de detectores de neutrinos para el monitoreo de reactores nucleares Uso de detectores de neutrinos para el monitoreo de reactores nucleares

Directory of Open Access Journals (Sweden)

Gerardo Moreno

2012-02-01

Full Text Available Se estudia la factibilidad del uso de los detectores de antineutrinos para el monitoreo de reactores nucleares. Usando un modelo sencillo de cascada de fisión a dos componentes, se ilustra la dependencia del número de antineutrinos detectados a una distancia L del reactor según la composición nuclear del combustible. Se explica el principio de detección de neutrinos de reactores en base al decaimiento beta inverso y se describe como los detectores de neutrinos pueden emplearse para el monitoreo de la producción de materiales fisibles en el reactor. Se comenta como generalizar este análisis al caso real de un reactor nuclear in situ y uno de los principales experimentos internacionales dedicados a este propósito. We study the feasibility to use antineutrinos detectors for monitoring of nuclear reactors. Using a simple model of fission shower with two components, we illustrate how the numbers of antineutrinos detected at a distance L from the reactor depend on the composition of the nuclear combustible. We explain the principles of reactor neutrino detection using inverse beta decays and we describe how neutrinos detectors can be used for monitoring the production of fissile materials within the reactors. We comment how to generalize this analysis to the realistic case of a nuclear reactor in situ and one of the main international experiments dedicated to study the use of neutrinos detectors as nuclear safeguards.

COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors

International Nuclear Information System (INIS)

Kaminaga, Masanori

1994-03-01

The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode as well as COOLOD-N code. In the COOLOD-N2 code, a 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. (author)

Stress distribution difference between Lava Ultimate full crowns and IPS e.max CAD full crowns on a natural tooth and on tooth-shaped implant abutments.

Science.gov (United States)

Krejci, Ivo; Daher, René

2017-04-01

The goal of this short communication is to present finite element analysis comparison of the stress distribution between CAD/CAM full crowns made of Lava Ultimate and of IPS e.max CAD, adhesively luted to natural teeth and to implant abutments with the shape of natural teeth. Six 3D models were prepared using a 3D content-creating software, based on a micro-CT scan of a human mandibular molar. The geometry of the full crown and of the abutment was the same for all models representing Lava Ultimate full crowns (L) and IPS e.max CAD full crowns (E) on three different abutments: prepared natural tooth (n), titanium abutment (t) and zirconia abutment (z). A static load of 400 N was applied on the vestibular and lingual cusps, and fixtures were applied to the base of the models. After running the static linear analysis, the post-processing data we analyzed. The stress values at the interface between the crown and the abutment of the Lt and Lz groups were significantly higher than the stress values at the same interface of all the other models. The high stress concentration in the adhesive at the interface between the crown and the abutment of the Lava Ultimate group on implants might be one of the factors contributing to the reported debondings of crowns.

Evaluation of neutron flux in the WWR-SM reactor channel and in the irradiating zone of U-150 cyclotron

International Nuclear Information System (INIS)

Sadikov, I.I.; Zinov'ev, V.G.; Sadikova, Z.O.; Salimov, M.I.

2006-01-01

Full text: For effective work of a reactor, and correct planning of experiments related to the reactor irradiation of various materials it is required to control a neutron flux in the given irradiation point for a long irradiation period. For realization of research works on topazes ennobling under irradiation by reactor neutrons as well as by secondary neutrons produced in a cyclotron it is necessary to know the total neutron flux and spectra. To resolve the problem a technique for registration of neutrons with different energy and calculation of a neutrons spectrum in the given irradiation points in reactor channels and in cyclotron behind the nickel target has been developed. Neutron flux density and energy spectra were monitored by use of the following nuclear reactions: 59 Co(n,γ) 60 Co, 197 Au(n,γ) 198 Au, 58 Ni(n,p) 58 Co, 24 Mg(n,p) 24 Na, 48 Ti(n,p) 48 Sc, 46 Ti(n,p) 46 Sc, 54 Fe(n,p) 54 Mn, 89 Y(n,2n) 88 Y, 60 Ni(np) 60 Co. Gamma spectrometer composed of HPGe detector (Rel. Eff. - 15%) and Digital Spectra Analyzer DSA-1000 (Canberra Ind., USA) was used to measure gamma activity of irradiated samples. Acquired gamma spectra were processed by means of Genie 2000 standard software package. The σ(E) functions and neutron spectra were calculated by using the least squares method and approximating the tabular and experimental data with power polynomials. The developed technique was applied for the adjustment of the topazes irradiation regimes in the reactor core and under secondary neutrons flux from a nickel target in the cyclotron. The given technique allows to calculate a logarithmic spectrum of neutrons in a energy range from 0,025 eV up to 12 MeV with the uncertainty of about 10 %. (author)

Simulación con el código MCNP del reactor nuclear RP-10 en su configuración #14, BOC

OpenAIRE

Lázaro, Gerardo; Parreño, Fernando

2001-01-01

Se presenta los resultados de exceso de reactividad del núcleo del reactor RP-10 en su configuración 14. Este exceso de reactividad ha sido calculado con MCNP4B con un modelo que describe en detalle las características de los elementos combustibles normales y de control, así como de cada elemento que constituye la configuración de trabajo #14. Este modelo fue previamente utilizado en el reactor RP-0 y ha sido aplicado en la configuración de arranque para el cálculo del exceso de reactividad y...

Damage by radiation in structural materials of BWR reactor vessels; Dano por radiacion en materiales estructurales de vasijas de reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

2002-07-01

The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA Mark III Salazar reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A{sup 2}). (Author)

Effect of temperature on two-phase anaerobic reactors treating slaughterhouse wastewater

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Simone Beux

2007-11-01

Full Text Available The effectiveness of the anaerobic treatment of effluent from a swine and bovine slaughterhouse was assessed in two sets of two-phase anaerobic digesters, operated with or without temperature control. Set A, consisting of an acidogenic reactor with recirculation and an upflow biological filter as the methanogenic phase, was operated at room temperature, while set B, consisting of an acidogenic reactor without recirculation and an upflow biological filter as the methanogenic phase, was maintained at 32°C. The methanogenic reactors showed COD (Chemical Demand of Oxygen removal above 60% for HRT (Hydraulic Retention Time values of 20, 15, 10, 8, 6, 4, and 2 days. When the HRT value in those reactors was changed to 1 day, the COD percentage removal decreased to 50%. The temperature variations did not have harmful effects on the performance of reactors in set A.Avaliou-se a eficiência do tratamento anaeróbio de efluente de matadouro de suínos e bovinos em dois conjuntos de biodigestores anaeróbios de duas fases, operados com e sem controle de temperatura. O conjunto A, formado por um reator acidogênico com recirculação e um filtro biológico de fluxo ascendente, foi operado a temperatura ambiente e o conjunto B, formado por um reator de fluxo ascendente e um filtro biológico de fluxo ascendente, foi mantido a 32°C. Os reatores metanogênicos apresentaram remoção de DQO acima de 60 % para os TRHs de 20, 15, 10, oito, seis, quatro e dois dias. Quando o TRH destes reatores foi mudado para um dia observou-se uma queda da porcentagem de remoção de DQO para 50 %. As variações de temperatura parecem não ter prejudicado o desempenho dos reatores do conjunto A.

Multi-purpose reactor

International Nuclear Information System (INIS)

1991-05-01

The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MW t h, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co 60 ) production capacity is 50000 Ci/yr, 200 Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N 16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

Modeling requirements for full-scope reactor simulators of fission-product transport during severe accidents

International Nuclear Information System (INIS)

Ellison, P.G.; Monson, P.R.; Mitchell, H.A.

1990-01-01

This paper describes in the needs and requirements to properly and efficiently model fission product transport on full scope reactor simulators. Current LWR simulators can be easily adapted to model severe accident phenomena and the transport of radionuclides. Once adapted these simulators can be used as a training tool during operator training exercises for training on severe accident guidelines, for training on containment venting procedures, or as training tool during site wide emergency training exercises

Procedures of ASME code case N-201 for KALIMER. Reactor internal structures

International Nuclear Information System (INIS)

Koo, Gyeong Hoi; Yoo, B.

2001-02-01

The main objective of this report is to describe the design procedure of ASME Boiler and Pressure Vessel Code, Code Case N-201-4, which is an elevated temperature structural design code of the Nuclear reactor internal structures, checking the criteria of stress limit, accumulated inelastic strain and deformation, creep-fatigue damage, and buckling limit. As one of examples, the creep-fatigue damage evaluations are carried out for the KALIMER reactor internal structures of baffle annulus. This report is expected to be very useful in evaluating the structural integrity of the liquid metal reactor operating under an elevated temperature

Inmigración e integración escolar en Baleares

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Carmen ORTE SOCÍAS

2006-01-01

Full Text Available En este artículo se analizan las dificultades del entorno social, político y económico en el que se inserta el sistema educativo en la Comunidad Autónoma de las Islas Baleares. Creemos imprescindible que el sistema educativo atienda a los colectivos recién llegados al mismo, de conformidad con los principios de interculturalidad e integración. Aún así, nos planteamos si es posible hacer frente a las problemáticas asociadas a la inmigración, exclusivamente a partir de las prácticas interculturales propiciadas por el sistema educativo.

Photocatalytic reactors for treating water pollution with solar illumination: a simplified analysis for n-steps flow reactors with recirculation

Energy Technology Data Exchange (ETDEWEB)

Sagawe, G.; Bahnemann, D. [Universitaet Hannover (Germany). Institut fuer Technische Chemie; Brandi, R.J.; Cassano, A.E. [INTEC Universidad Nacional del Litoral and CONICET, Sante Fe (Argentina)

2005-09-01

The concentration of dissolved oxygen in water, in equilibrium with atmospheric air (ca. 8 ppm at 20{sup o}C), defines the limits of all practical oxidizing processes for removing pollutants in photocatalytic reactors. To solve this limitation, an alternative approach to that of a continuously aerated reactor is the use of a recirculating system with aeration performed after every cycle at the reactor entering stream. As defined by the nature of a single recirculating step (the need of a reactor operation at a rather low concentration range), this procedure results in a very low photonic efficiency (thus requiring a large photon collecting area and consequently increasing the capital cost). The design engineer will have to resort to a series of several reactors with recirculation. This solution may then lead to a very high Photonic Efficiency for the entire process (i.e., a reduced light harvesting area) at the price of an increase in the required capital cost (due to the larger number of reactors). This paper provides a very simple analysis and analytical expressions that can be used to estimate, for a desired degree of degradation, a trade-off solution between a high number of reactors and a very large surface area to collect the solar photons. (author)

The 4th surveillance testing for Kori unit 3 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

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

2000-10-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

Development of a full scope reactor engineering simulator

International Nuclear Information System (INIS)

Venhuizen, J.R.; Laats, E.T.

1988-01-01

An engineering laboratory is pursuing the development of an engineering simulator for use by several agencies of the U.S. Government. According to the authors, this simulator will provide the highest fidelity simulation with initial objectives for studying augmented nuclear reactor operator training, and later for advanced concepts testing as applicable to control room accident diagnosis and management

Investigación formativa e investigación productiva de conocimiento en la universidad

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Bernardo Restrepo Gómez

2003-04-01

Full Text Available Este artículo discute sobre dos objetivos que tiene la universidad frente a la investigación, a saber: enseñar a investigar y hacer investigación. Presenta diferentes acepciones del constructo “Investigación Formativa”, así como formas de implementación de ésta. Finalmente, distingue entre investigación formativa e investigación en sentido estricto.

Impact of partial nitritation degree and C/N ratio on simultaneous Sludge Fermentation, Denitrification and Anammox process.

Science.gov (United States)

Wang, Bo; Peng, Yongzhen; Guo, Yuanyuan; Yuan, Yue; Zhao, Mengyue; Wang, Shuying

2016-11-01

This study presents a novel process (i.e. PN/SFDA) to remove nitrogen from low C/N domestic wastewater. The process mainly involves two reactors, a pre-Sequencing Batch Reactor for partial nitritation (termed as PN-SBR) and an anoxic reactor for integrated Denitrification and Anammox with carbon sources produced from Sludge Fermentation (termed as SFDA). During long-term Runs, NO2(-)/NH4(+) ratio (i.e. NO2(-)-N/NH4(+)-N calculated by mole) in the PN-SBR effluent was gradually increased from 0.2 to 37 by extending aerobic duration, meaning that partial nitritation turning to full nitritation could be achieved. Impact of partial nitritation degree on SFDA process was investigated and the result showed that, NO2(-)/NH4(+) ratios between 2 and 10 were appropriate for the co-existence of denitrification and anammox together in the SFDA reactor, and denitrification instead of anammox contributed greater for nitrogen removal. Further batch tests indicated that anammox collaborated well with denitrification at low C/N (1.0 in this study). Copyright © 2016 Elsevier Ltd. All rights reserved.

Presentación. Globalización e inmigración:los debates actuales.

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Carlota Solé

2006-01-01

Full Text Available Este artículo sirve de presentación al monográfico de la Revista Española de Investigaciones Sociológicas sobre "Globalización e inmigración". Comienza poniendo de relieve algunos de los cambios que se están produciendo en los flujos migratorios (como su feminización y en las políticas de regulación de las migraciones (como las políticas europeas. A continuación apunta brevemente algunos de los debates teóricos y políticos actuales en torno a la inmigración en el marco de la globalización: transnacionalidad, ciudadanía e integración, identidad europea y multiculturalidad. La segunda parte del artículo presenta el desarrollo académico de las teorías de las migraciones a través de la comparación del contenido del Handbook of International Migration de 1999 y los artículos recogidos en el monográfico de la International Migration Review de 2004. Por último, sintetiza los artículos y notas de investigación de este monográfico de la Revista Española de Investigaciones Sociológicas, ubicándolos en esos desarrollos teóricos.

A computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Hu, Rui, E-mail: rhu@anl.gov; Yu, Yiqi

2016-11-15

Highlights: • Developed a computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors. • Applied fully-coupled JFNK solution scheme to avoid the operator-splitting errors. • The accuracy and efficiency of the method is confirmed with a 7-assembly test problem. • The effects of different spatial discretization schemes are investigated and compared to the RANS-based CFD simulations. - Abstract: For efficient and accurate temperature predictions of sodium fast reactor structures, a 3-D full-core conjugate heat transfer modeling capability is developed for an advanced system analysis tool, SAM. The hexagon lattice core is modeled with 1-D parallel channels representing the subassembly flow, and 2-D duct walls and inter-assembly gaps. The six sides of the hexagon duct wall and near-wall coolant region are modeled separately to account for different temperatures and heat transfer between coolant flow and each side of the duct wall. The Jacobian Free Newton Krylov (JFNK) solution method is applied to solve the fluid and solid field simultaneously in a fully coupled fashion. The 3-D full-core conjugate heat transfer modeling capability in SAM has been demonstrated by a verification test problem with 7 fuel assemblies in a hexagon lattice layout. Additionally, the SAM simulation results are compared with RANS-based CFD simulations. Very good agreements have been achieved between the results of the two approaches.

Efecto de la concentración de SO2 en las reacciones de calcinación y sulfatación de calcáreos en reactores de lecho fluidizado. // Effect of the SO2 concentration in the calcinations and sulfatation reactions in a fluidized bed reactor.

Directory of Open Access Journals (Sweden)

J. E. Lindo Samaniego

2008-01-01

Full Text Available Fue realizado un estudio sobre el efecto de las concentraciones de SO2 en la absorción por calcáreo en hornos de LechoFluidizado. Para observar la influencia del SO2 en los parámetros de diferentes procesos físicos y químicos fueron creadosambientes para cuatro concentraciones diferentes de SO2 : 500, 1000, 2000 y 4000 ppm. Se utilizaron dos tipos decalcáreos: Dolimitico-DP y el Calcítico-CI. El Lecho Fluidizado Burbujeante utilizado tiene 160 mm de diámetro interno yfue fluidizado con aire a la temperatura de 850 °C, con una concentración de SO2 deseada. Como material del lecho fueutilizada la arena de cuarzo (99,9% con diámetro de 385 μm y de masa aproximadamente 3,0 kg. El calcáreo fueadicionado en dosificación de 50 g con el reactor ya pre-calentado. Las variaciones de las concentraciones de SO2, CO2,CO, O2 y las descargas, fueron monitoreadas continuamente a la salida del ciclón que fue utilizado para la retención de lapartícula fina. Para esos dados se desarrolló un programa en LabView. El modelo matemático escogido posibilitó ladeterminación de la conversión.Palabras claves: Lecho fluidizado, dióxido de azufre, absorción de azufre, calcáreo , reactor de lechofluidizado.___________________________________________________________________________Abstract.A Study of the effect of the concentrations of SO2 in its absorption by limestones in fluidized bed furnaces wasconducted. For the determination of the SO2 influence on the different physical and chemical parameters of process,such as calcinations and sulfatation four different atmospheres were used in the reator with concentrations of SO2 of500, 1000, 2000 and 4000 ppm. Two types of limestones were used: Dolomite-DP and Calcitic-CI. The bench scalebubbling fluidized bed reactor had a 160 mm internal diameter and was fluidized with air at 850 °C containing therequired concentration of SO2. Bed material was quartz sand (99,9% , with 385 μm diameter and approximately

Ética e investigación

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Pedro Alvarez Viera

2018-02-01

Full Text Available Iniciamos esta cartilla para el CEIDE teniendo en cuenta la guía de opciones de grado para la facultad de Derecho, CEIDE, 2017(1, y el libro del profesor Pedro González.(2. Recomiendo tener claridad en los siguientes temas. Curso de argumentación jurídica de Manuel Atienza.(3. Ante el monopolio de lo científico propongo la ética del discurso, de Apel, y la ética material, de Enrique Dussel.(4, Para centrar los temas a trabajar desde el derecho funcional como también desde lo justo y no solo desde la estructura de la norma, el derecho positivo. En la primera parte se trabaja el currículo de capacitación sobre ética e investigación con sus cuatro secciones, y un módulo de ética aplicada a la investigación. En la segunda parte vemos los aspectos generales sobre el consentimiento informado, así como la metodología de la ética e investigación. En la tercera parte están los documentos más importantes como guía para hacer investigación, así mismo la bibliografía.

HECTR [Hydrogen Event Containment Transient Response] Version 1.5N: A modification of HECTR Version 1.5 for application to N Reactor

International Nuclear Information System (INIS)

Camp, A.L.; Dingman, S.E.

1987-05-01

This report describes HECTR Version 1.5N, which is a special version of HECTR developed specifically for application to the N Reactor. HECTR is a fast-running, lumped-parameter containment analysis computer program that is most useful for performing parametric studies. The main purpose of HECTR is to analyze nuclear reactor accidents involving the transport and combustion of hydrogen, but HECTR can also function as an experiment analysis tool and can solve a limited set of other types of containment problems. Version 1.5N is a modification of Version 1.5 and includes changes to the spray actuation logic, and models for steam vents, vacuum breakers, and building cross-vents. Thus, all of the key features of the N Reactor confinement can be modeled. HECTR is designed for flexibility and provides for user control of many important parameters, if built-in correlations and default values are not desired

Delayed photoneutrons of the of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Ngo Quang Huy; Ha Van Thong; Vu Hai Long; Ngo Phu Khang; Nguyen Nhi Dien; Pham Van Lam; Huynh Dong Phuong; Luong Ba Vien; Le Vinh Vinh

1994-01-01

Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10 5 /10 8 n/cm 2 /sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to (γ,n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is β B e eff =0.49%β eff for a beryllium weight relative to U 235 fuel of m B e/m U = 8.5. This result is acceptable in comparison to those obtained for other Be-U 235 media. (author). 5 refs., 2 figs., 4 tabs

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

International Nuclear Information System (INIS)

Sri Kuntjoro; Pande Made Udiyani

2016-01-01

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

Kinetic modelling and characterization of microbial community present in a full-scale UASB reactor treating brewery effluent.

Science.gov (United States)

Enitan, Abimbola M; Kumari, Sheena; Swalaha, Feroz M; Adeyemo, J; Ramdhani, Nishani; Bux, Faizal

2014-02-01

The performance of a full-scale upflow anaerobic sludge blanket (UASB) reactor treating brewery wastewater was investigated by microbial analysis and kinetic modelling. The microbial community present in the granular sludge was detected using fluorescent in situ hybridization (FISH) and further confirmed using polymerase chain reaction. A group of 16S rRNA based fluorescent probes and primers targeting Archaea and Eubacteria were selected for microbial analysis. FISH results indicated the presence and dominance of a significant amount of Eubacteria and diverse group of methanogenic Archaea belonging to the order Methanococcales, Methanobacteriales, and Methanomicrobiales within in the UASB reactor. The influent brewery wastewater had a relatively high amount of volatile fatty acids chemical oxygen demand (COD), 2005 mg/l and the final COD concentration of the reactor was 457 mg/l. The biogas analysis showed 60-69% of methane, confirming the presence and activities of methanogens within the reactor. Biokinetics of the degradable organic substrate present in the brewery wastewater was further explored using Stover and Kincannon kinetic model, with the aim of predicting the final effluent quality. The maximum utilization rate constant U max and the saturation constant (K(B)) in the model were estimated as 18.51 and 13.64 g/l/day, respectively. The model showed an excellent fit between the predicted and the observed effluent COD concentrations. Applicability of this model to predict the effluent quality of the UASB reactor treating brewery wastewater was evident from the regression analysis (R(2) = 0.957) which could be used for optimizing the reactor performance.

Dispositivo de posicionamiento de muestras biológicas para su irradiación en un canal radial de un reactor nuclear // Biological samples positioning device for irradiations on a radial channel at the nuclear research reactor

Directory of Open Access Journals (Sweden)

Maritza Rodríguez - Gual

2010-05-01

Full Text Available ResumenPor la demanda de un dispositivo experimental para el posicionamiento de las muestras biológicaspara su irradiación en un canal radial de un reactor nuclear de investigaciones en funcionamiento, seconstruyó y se puso en marcha un dispositivo para la colocación y retirada de las muestras en laposición de irradiación de dicho canal. Se efectuaron las valoraciones económicas comparando conotro tipo de dispositivo con las mismas funciones. Este trabajo formó parte de un proyectointernacional entre Cuba y Brasil que abarcó el estudio de los daños inducidos por diferentes tipos deradiación ionizante en moléculas de ADN. La solución propuesta es comprobada experimentalmente,lo que demuestra la validez práctica del dispositivo. Como resultado del trabajo, el dispositivoexperimental para la irradiación de las muestras biológicas se encuentra instalado y funcionando yapor 5 años en el canal radial # 3(BH#3 Palabras claves: reactor nuclear de investigaciones, dispositivo para posicionamiento de muestras,___________________________________________________________________________AbstractFor the demand of an experimental device for biological samples positioning system for irradiationson a radial channel at the nuclear research reactor in operation was constructed and started up adevice for the place and remove of the biological samples from the irradiation channels withoutinterrupting the operation of the reactor. The economical valuations are effected comparing withanother type of device with the same functions. This work formed part of an international projectbetween Cuba and Brazil that undertook the study of the induced damages by various types ofionizing radiation in DNA molecules. Was experimentally tested the proposed solution, whichdemonstrates the practical validity of the device. As a result of the work, the experimental device forbiological samples irradiations are installed and operating in the radial beam hole #3(BH#3

Full-Scale Modeling Explaining Large Spatial Variations of Nitrous Oxide Fluxes in a Step-Feed Plug-Flow Wastewater Treatment Reactor.

Science.gov (United States)

Ni, Bing-Jie; Pan, Yuting; van den Akker, Ben; Ye, Liu; Yuan, Zhiguo

2015-08-04

Nitrous oxide (N2O) emission data collected from wastewater treatment plants (WWTPs) show huge variations between plants and within one plant (both spatially and temporarily). Such variations and the relative contributions of various N2O production pathways are not fully understood. This study applied a previously established N2O model incorporating two currently known N2O production pathways by ammonia-oxidizing bacteria (AOB) (namely the AOB denitrification and the hydroxylamine pathways) and the N2O production pathway by heterotrophic denitrifiers to describe and provide insights into the large spatial variations of N2O fluxes in a step-feed full-scale activated sludge plant. The model was calibrated and validated by comparing simulation results with 40 days of N2O emission monitoring data as well as other water quality parameters from the plant. The model demonstrated that the relatively high biomass specific nitrogen loading rate in the Second Step of the reactor was responsible for the much higher N2O fluxes from this section. The results further revealed the AOB denitrification pathway decreased and the NH2OH oxidation pathway increased along the path of both Steps due to the increasing dissolved oxygen concentration. The overall N2O emission from this step-feed WWTP would be largely mitigated if 30% of the returned sludge were returned to the Second Step to reduce its biomass nitrogen loading rate.

Color coherent effects in (e,e{prime}N) and (e,e{prime}N,N(h)) processes at CEBAF

Energy Technology Data Exchange (ETDEWEB)

Frankfurt, L.L.; Sargsyan, M.M. [Tel Aviv Univ. (Israel); Strikman, M.I. [Pennsylvania State Univ., University Park, PA (United States)]|[St. Petersburg Nuclear Physics Inst. (Russian Federation)

1994-04-01

The options for investigating color coherent effects and competing nuclear effects of nucleon-nucleon correlations in nuclei, nuclear shell effects in (e, e{prime}N) and (e, e{prime}NN(h)) reactions are considered. They argue that extension of CEBAF energies to reach Q{sup 2} = 10 GeV{sup 2} will allow systematical investigations of color coherent effects in nonperturbative regime of QCD and their interplay with nuclear effects.

Studies of fragileness in steels of vessels of BWR reactors; Estudios de fragilizacion en aceros de vasija de reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2003-07-01

The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA MARK lll reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A{sup 2}. (Author)

Reactor Dosimetry State of the Art 2008

Science.gov (United States)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

data, damage correlations. Two-dimensional mapping of the calculated fission power for the full-size fuel plate experiment irradiated in the advanced test reactor / G. S. Chang and M. A. Lillo. The radiation safety information computational center: a resource for reactor dosimetry software and nuclear data / B. L. Kirk. Irradiated xenon isotopic ratio measurement for failed fuel detection and location in fast reactor / C. Ito, T. Iguchi and H. Harano. Characterization of dosimetry of the BMRR horizontal thimble tubes and broad beam facility / J.-P. Hu, R. N. Reciniello and N. E. Holden. 2007 nuclear data review / N. E. Holden. Further dosimetry studies at the Rhode Island nuclear science / R. N. Reciniello ... [et al.]. Characterization of neutron fields in the experimental fast reactor Joyo MK-III core / S. Maeda ... [et al.]. Measuring [symbol]Li(n, t) and [symbol]B(n, [symbol]) cross sections using the NIST alpha-gamma apparatus / M. S. Dewey ... [et al.]. Improvement of neutron/gamma field evaluation for restart of JMTR / Y. Nagao ... [et al.]. Monitoring of the irradiated neutron fluence in the neutron transmutation doping process of HANARO / M.-S. Kim and S.-J. Park.Training reactor VR-l neutron spectrum determination / M. Vins, A. Kolros and K. Katovsky. Differential cross sections for gamma-ray production by 14 MeV neutrons on iron and bismuth / V. M. Bondar ... [et al.]. The measurements of the differential elastic neutron cross-sections of carbon for energies from 2 to 133 ke V / O. Gritzay ... [et al.]. Determination of neutron spectrum by the dosimetry foil method up to 35 Me V / S. P. Simakov ... [et al.]. Extension of the BGL broad group cross section library / D. Kirilova, S. Belousov and Kr. Ilieva. Measurements of neutron capture cross-section for tantalum at the neutron filtered beams / O. Gritzayand V. Libman. Measurements of microscopic data at GELINA in support of dosimetry / S. Kopecky ... [et al.]. Nuclide guide and international chart of

Analysis Of Core Management For The Transition Cores Of RSG-GAS Reactor To Full-Silicide Core

International Nuclear Information System (INIS)

Malem Sembiring, Tagor; Suparlina, Lily; Tukiran

2001-01-01

The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 g U/cc is still doing. At the end of 2000, the reactor has been operated for 3 transition cores which is the mixed core of oxide-silicide. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 10 transition cores to achieve a full-silicide core. The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters such as excess reactivity and shutdown margin. The measurement of the core parameters was carried out using the method of compensation of couple control rods. The experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safety to a full-silicide core

Strategy to identify the causes and to solve a sludge granulation problem in methanogenic reactors: application to a full-scale plant treating cheese wastewater.

Science.gov (United States)

Macarie, Hervé; Esquivel, Maricela; Laguna, Acela; Baron, Olivier; El Mamouni, Rachid; Guiot, Serge R; Monroy, Oscar

2017-08-26

Granulation of biomass is at the basis of the operation of the most successful anaerobic systems (UASB, EGSB and IC reactors) applied worldwide for wastewater treatment. Despite of decades of studies of the biomass granulation process, it is still not fully understood and controlled. "Degranulation/lack of granulation" is a problem that occurs sometimes in anaerobic systems resulting often in heavy loss of biomass and poor treatment efficiencies or even complete reactor failure. Such a problem occurred in Mexico in two full-scale UASB reactors treating cheese wastewater. A close follow-up of the plant was performed to try to identify the factors responsible for the phenomenon. Basically, the list of possible causes to a granulation problem that were investigated can be classified amongst nutritional, i.e. related to wastewater composition (e.g. deficiency or excess of macronutrients or micronutrients, too high COD proportion due to proteins or volatile fatty acids, high ammonium, sulphate or fat concentrations), operational (excessive loading rate, sub- or over-optimal water upflow velocity) and structural (poor hydraulic design of the plant). Despite of an intensive search, the causes of the granulation problems could not be identified. The present case remains however an example of the strategy that must be followed to identify these causes and could be used as a guide for plant operators or consultants who are confronted with a similar situation independently of the type of wastewater. According to a large literature based on successful experiments at lab scale, an attempt to artificially granulate the industrial reactor biomass through the dosage of a cationic polymer was also tested but equally failed. Instead of promoting granulation, the dosage caused a heavy sludge flotation. This shows that the scaling of such a procedure from lab to real scale cannot be advised right away unless its operability at such a scale can be demonstrated.

Neutron flux measurements in PUSPATI Triga Reactor

International Nuclear Information System (INIS)

Gui Ah Auu; Mohamad Amin Sharifuldin Salleh; Mohamad Ali Sufi.

1983-01-01

Neutron flux measurement in the PUSPATI TRIGA Reactor (PTR) was initiated after its commissioning on 28 June 1982. Initial measured thermal neutron flux at the bottom of the rotary specimen rack (rotating) and in-core pneumatic terminus were 3.81E+11 n/cm 2 sec and 1.10E+12n/cm 2 sec respectively at 100KW. Work to complete the neutron flux data are still going on. The cadmium ratio, thermal and epithermal neutron flux are measured in the reactor core, rotary specimen rack, in-core pneumatic terminus and thermal column. Bare and Cadmium covered gold foils and wires are used for the above measurement. The activities of the irradiated gold foils and wires are determined using Ge(Li) and hyperpure germinium detectors. (author)

The Jules Horowitz reactor project, a driver for revival of the research reactor community

Energy Technology Data Exchange (ETDEWEB)

Pere, P.; Cavailler, C.; Pascal, C. [AREVA TA, CEA Cadarache - Etablissement d' AREVA TA - Chantier RJH - MOE - BV2 - BP no. 9 - 13115 Saint Paul lez Durance (France); CS 50497 - 1100, rue JR Gauthier de la Lauziere, 13593 Aix en Provence cedex 3 (France)

2010-07-01

The first concrete of the nuclear island for the Jules Horowitz Reactor (JHR) was poured at the end of July 2009 and construction is ongoing. The JHR is the largest new platform for irradiation experiments supporting Generation II and III reactors, Generation IV technologies, and radioisotope production. This facility, composed of a unique grouping of workshops, hot cells and hot laboratories together with a first -rate MTR research reactor, will ensure that the process, from preparations for irradiation experiments through post-irradiation non-destructive examination, is completed expediently, efficiently and, of course, safely. In addition to the performance requirements to be met in terms of neutron fluxes on the samples (5x10{sup 14} n.cm{sup -2}/sec{sup -1} E> 1 MeV in core and 3,6x10{sup 14} n.cm{sup -2}/sec{sup -1} E<0.625 eV in the reflector) and the JHR's considerable irradiation capabilities (more than 20 experiments and one-tenth of irradiation area for simultaneous radioisotope production), the JHR is the first MTR to be built since the end of the 1960's, making this an especially challenging project. The presentation will provide an overview of the reactor, hot cells and laboratories and an outline of the key milestones in the project schedule, including initial criticality in early 2014 and radioisotope production in 2015. This will be followed by a description of the project organization set up by the CEA as owner and future operator and AREVA TA as prime contractor and supplier of critical systems, and a discussion of project challenges, especially those dealing with the following items: - accommodation of a broad experimental domain, - involvement by international partners making in-kind contributions to the project, - development of components critical to safety and performance, - the revival of engineering of research reactors and experimental devices involving France's historical players in the field of research reactors, and

k0-measurements and related nuclear data compilation for (n, γ) reactor neutron activation analysis Pt. 3b

International Nuclear Information System (INIS)

Corte, F. de; Simonits, A.

1989-01-01

k 0 -factors and related nuclear data are tabulated for 112 radionuclides of interest in (n, γ) reactor neutron activation analysis. Whenever relevant, critical comments are made with respect to the accuracy of literature data for e. g. isotopic abundances, half-lives, absolute gamma-intensities and 2200 m · s -1 (n, γ) cross sections. As to the latter, a comparison is made with the values calculated from the experimentally determined k 0 -factors, by introduction of selected literature data for the input parameters. References to the table (79 pages) include 156 items. (author) 7 refs.; 1 tab

Prefermentación de agua residual urbana empleando un reactor biopelícula de lecho sumergido

OpenAIRE

Castillo de Castro, Pedro; González Ruiz, Guarocuya; Tejero Monzón, Iñaki

1999-01-01

Esta investigación describe la acidificación de agua residual urbana (ARU). La experimentación se llevó a cabo utilizando un reactor biopelícula de lecho sumergido fijo (BLSF), en el cual se produce la hidrólisis de los sólidos en suspensión retenidos en la zona de decantación del reactor, mientras que la biomasa de la biopelícula se encarga de la acidificación del medio. Se operó bajo cargas orgánicas aplicadas (COA) comprendidas entre 0,5 y 23 kg DQO/m3·d, tiempos de retención hidráulico (T...

REMOCIÓN DE ARSÉNICO ASISTIDA POR OXIDACIÓN UV SOLAR (RAOS EN FOTO-REACTORES TUBULARES DE SECCIÓN SEMICIRCULAR - CINÉTICA DEL CRECIMIENTO DE FLÓCULOS DE Fe(OH3

Directory of Open Access Journals (Sweden)

Ramiro Escalera Vásquez

2012-01-01

Cálculos de capacidad de tratamiento, en régimen continuo (considerando tiempos de residencia hidráulica iguales a los tiempos de irradiación, demuestran la mayor capacidad del foto-reactor de 71 cm de diámetro, logrando un flujo diario de 190 Lm-2 para una operación de 5 h por día. Desde el punto de vista económico y de su construcción, este foto-reactor es más práctico que los reactores de menor diámetro, por la menor cantidad de accesorios y materiales involucrados.

Determination of the lowest critical power levels of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Binh, Do Quang; Nghiem, Huynh Ton; Tuan, Nguyen Minh; Vien, Luong Ba; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

This paper presents the experimental methods for determining critical states of the Dalat Nuclear Research Reactor containing an extraneous neutron source induced by gamma ray reactions on beryllium in the reactor. The lowest critical power levels are measured at various moments after the reactor is shut down following 100 hours of its continuous operation. Th power levels vary from (0.5-1.2) x 10{sup -4} of P{sub n}, i.e. (25-60)W to (1.1-1.6) x 10{sup -5} of P{sub n}, i.e. (5.5-8)W at corresponding times of 4 days to 13 days after the reactor is shut down. However the critical power must be chosen greater than 500 W to sustain the steady criticality of the reactor for a long time. (author). 3 refs. 4 figs. 1 tab.

Intact and Degraded Component Criticality Calculations of N Reactor Spent Nuclear Fuel

International Nuclear Information System (INIS)

L. Angers

2001-01-01

The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k eff ) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR

Delayed photoneutrons of the of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Khang, Ngo Phu; Dien, Nguyen Nhi; Lam, Pham Van; Phuong, Huynh Dong; Vien, Luong Ba; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10{sup 5}/10{sup 8} n/cm{sup 2}/sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to ({gamma},n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is {beta}{sup B}e{sub eff}=0.49%{beta}{sub eff} for a beryllium weight relative to U{sup 235} fuel of m{sub B}e/m{sub U} = 8.5. This result is acceptable in comparison to those obtained for other Be-U{sup 235} media. (author). 5 refs., 2 figs., 4 tabs.

La Cooperación Internacional para la formación e inserción sociolaboral de jóvenes mexicanos

Directory of Open Access Journals (Sweden)

Cristina Girardo

2012-01-01

Full Text Available Hemos estructurado este trabajo en torno a la actuación de la Cooperación Internacional y su vínculo con organizaciones de la sociedad civil para responder a la problemática particular de la formación e inserción sociolaboral de los jóvenes y las jóvenes de México. Planteamos cuestiones vinculadas al tema de estudio como, por ejemplo, la estructuración de proyectos de la sociedad civil vinculados a la formación e inserción sociolaboral juvenil en México, y las propuestas desde las agencias de la cooperación internacional; así como también analizamos los procedimientos de vinculación estratégica entre los distintos actores que participan en estos procesos que se organizan para promover tendencias de formación e inserción sociolaboral juvenil en México.

Investigación e investigación formativa

OpenAIRE

Carlos Augusto Hernández

2003-01-01

Este artículo hace una distinción entre investigación e investigación formativa. También señala los peligros de confundir estas dos formas de trabajo académico e insiste en la necesidad de desarrollarlas simultáneamente para enfrentar los nuevos retos de la educación superior. Igualmente habla de la importancia de intensificar los vínculos entre investigación y docencia y hace la diferencia entre investigación pedagógica e investigación formativa. Finalmente cuestiona la posibilidad de una fo...

Efecto de la concentración de SO2 en las reacciones de calcinación y sulfatación de calcáreos en reactores de lecho fluidizado.

Directory of Open Access Journals (Sweden)

J. E. Lindo Samaniego

2008-01-01

Full Text Available Fue realizado un estudio sobre el efecto de las concentraciones de SO2 en la absorción por calcáreo en hornos de Lecho Fluidizado. Para observar la influencia del SO2 en los parámetros de diferentes procesos físicos y químicos fueron creados ambientes para cuatro concentraciones diferentes de SO2 : 500, 1000, 2000 y 4000 ppm. Se utilizaron dos tipos de calcáreos: Dolimitico-DP y el Calcítico-CI. El Lecho Fluidizado Burbujeante utilizado tiene 160 mm de diámetro interno y fue fluidizado con aire a la temperatura de 850 °C, con una concentración de SO2 deseada. Como material del lecho fue utilizada la arena de cuarzo (99,9% con diámetro de 385 µm y de masa aproximadamente 3,0 kg. El calcáreo fue adicionado en dosificación de 50 g con el reactor ya pre-calentado. Las variaciones de las concentraciones de SO2, CO2, CO, O2 y las descargas, fueron monitoreadas continuamente a la salida del ciclón que fue utilizado para la retención de la partícula fina. Para esos dados se desarrolló un programa en LabView. El modelo matemático escogido posibilitó la determinación de la conversión.A study of the effect of the concentrations of SO2 in its absorption by limestones in fluidized bed furnaces was conducted. For the determination of the SO2 influence on the different physical and chemical parameters of process, such as calcinations and sulfatation four different atmospheres were used in the reator with concentrations of SO2 of 500, 1000, 2000 and 4000 ppm. Two types of limestones were used: Dolomite-DP and Calcitic-CI. The bench scale bubbling fluidized bed reactor had a 160 mm internal diameter and was fluidized with air at 850 °C containing the required concentration of SO2. Bed material was quartz sand (99,9% , with 385 µm diameter and approximately 3 kg of mass. The limestone was introduced in samples of 50 kg in the reactor previously stabilized and their discharges in the reactor exit were continually monitored. For the

Caracterización mecánica y microestructural de aceros de baja activación candidatos como primera pared en los reactores de fusión por confinamiento magnético

Directory of Open Access Journals (Sweden)

Hernández, M. T.

1996-04-01

Full Text Available Currently, the design development of fusion reactors and the possible materials to use in them are being studied in parallel. One of the most critical problems in this research is the structural material selection for the first wall and blanket. The aim of the present work is to study three low activation alloys designed in Germany in which niobium has been substituted by tantalum or cerium. The mechanical results show that the alloys containing cerium are in the same order of the low activation materials known to date, but the tantalum doped alloy produces TaC₃ precipitation that destabilizes the matrix and provokes large microstructural changes. This causes a decrease of the mechanical properties at about 600 °C. This fact makes this alloy unsuitable for the first wall on fusion reactors, because the working temperature is near 550 °C.

Actualmente, se está estudiando de forma paralela al desarrollo del diseño de los reactores de fusión los posibles materiales a emplear en estos. Una de las cuestiones más críticas en esta investigación es la selección del material estructural a emplear como primera pared y envoltura. En el presente trabajo se estudian tres aleaciones de diseño alemán de baja activación, en las que se ha sustituido el niobio por tantalio o por cerio. Los resultados mecánicos muestran que las aleaciones, en las que se ha añadido cerio de forma controlada, están en la línea de las de baja activación existentes hasta ahora, pero la adición de tantalio presenta problemas al provocar una precipitación primaria y masiva de carburos TaC₃ que desestabiliza la matriz y origina cambios microestructurales muy acusados. Así, ocurre un descenso en las propiedades mecánicas en torno a los 600 °C, que la incapacita para la aplicación como primera pared, ya que la temperatura de trabajo de ésta se halla en torno a los 550 °C.

Información e Inteligencia: una reflexión interdisciplinar

Directory of Open Access Journals (Sweden)

Álvaro Cremades y Gustavo Díaz Matey

2015-12-01

Full Text Available La información es la materia prima con la que se construye la inteligencia. En consecuencia, la identificación, definición y clasificación de fuentes que provean de información veraz y pertinente representa una de las más elementales tareas en el proceso de producción de inteligencia, sea cual fuere la forma que este adopte. Sin embargo, el desigual e insuficiente desarrollo teórico de los estudios de inteligencia ha supuesto un grave impedimento para la clarificación de estos aspectos esenciales en torno al concepto de información, habiéndose afrontado hasta el momento y salvo honrosas pero escasas excepciones de forma meramente empírica o parcial basándose en análisis casuísticos derivados de evoluciones históricas o procedimientos y prácticas llevadas a cabo en distintas estructuras.Information and intelligence an interdisciplinary reflection

Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This guide discusses test procedures that can be used in conjunction with, but not as alternatives to, those required by Practices E185 and E2215 for the surveillance of nuclear reactor vessels. The supplemental mechanical property tests outlined permit the acquisition of additional information on radiation-induced changes in fracture toughness, notch ductility, and yield strength properties of the reactor vessel steels. 1.2 This guide provides recommendations for the preparation of test specimens for irradiation, and identifies special precautions and requirements for reactor surveillance operations and postirradiation test planning. Guidance on data reduction and computational procedures is also given. Reference is made to other ASTM test methods for the physical conduct of specimen tests and for raw data acquisition.

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

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

Measurement of 89Y(n,2n) spectral averaged cross section in LR-0 special core reactor spectrum

Science.gov (United States)

Košťál, Michal; Losa, Evžen; Baroň, Petr; Šolc, Jaroslav; Švadlenková, Marie; Koleška, Michal; Mareček, Martin; Uhlíř, Jan

2017-12-01

The present paper describes reaction rate measurement of 89Y(n,2n)88Y in a well-defined reactor spectrum of a special core assembled in the LR-0 reactor and compares this value with results of simulation. The reaction rate is derived from the measurement of activity of 88Y using gamma-ray spectrometry of irradiated Y2O3 sample. The resulting cross section value averaged in spectrum is 43.9 ± 1.5 μb, averaged in the 235U spectrum is 0.172 ± 0.006 mb. This cross-section is important as it is used as high energy neutron monitor and is therefore included in the International Reactor Dosimetry and Fusion File. Calculations of reaction rates were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF nuclear data libraries. The agreement with uranium description by CIELO library is very good, while in ENDF/B-VII.0 description of uranium, underprediction about 10% in average can be observed.

Seismic responses of N-Reactor core. Independent review of Phase II work

International Nuclear Information System (INIS)

Chen, J.C.; Lo, T.; Chinn, D.J.; Murray, R.C.; Johnson, J.J.; Maslenikov, O.R.

1985-08-01

Seismic response of the N-Reactor core was independently analyzed to validate the results of Impell's analysis. The analysis procedure consists of two major stages: linear soil-structure interaction (SSI) analysis of the overall N-Reactor structure complex and nonlinear dynamic analysis of the reactor core. In the SSI analysis, CLASSI computer codes were used to calculate the SSI response of the structures and to generate the input motions for the nonlinear reactor core analysis. In addition, the response was compared to the response from the SASSI analysis under review. The impact of foundation modeling techniques and the effect of soil stiffness variation on SSI response were also investigated. In the core analysis, a nonlinear dynamic analysis model was developed. The stiffness representation of the model was calculated through a finite element analysis of several local core geometries. Finite element analyses were also used to study the block to block interaction characteristics. Using this nonlinear dynamic model along with the basemat time histories generated from CLASSI and SASSI, several dynamic analyses of the core were performed. A series of sensitivity studies was performed to investigate the discretization of the core, the effect of vertical acceleration, the effect of basemat rocking, and modeling assumptions. In general, our independent analysis of core response validates the order of magnitude of the displacement calculated by Impell. 11 refs., 110 figs., 12 tabs

Damage by radiation in structural materials of BWR reactor vessels

International Nuclear Information System (INIS)

Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

2002-01-01

The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA Mark III Salazar reactor and separately with Ni +3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A 2 ). (Author)

Studies of fragileness in steels of vessels of BWR reactors

International Nuclear Information System (INIS)

Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

2003-01-01

The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA MARK lll reactor and separately with Ni +3 ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A 2 . (Author)

arXiv Performance of a full scale prototype detector at the BR2 reactor for the SoLid experiment

CERN Document Server

Abreu, Y.; Arnold, L.; Ban, G.; Beaumont, W.; Bongrand, M.; Boursette, D.; Castle, B.C.; Clark, K.; Coupé, B.; Cussans, D.; De Roeck, A.; D'Hondt, J.; Durand, D.; Fallot, M.; Ghys, L.; Giot, L.; Guillon, B.; Ihantola, S.; Janssen, X.; Kalcheva, S.; Kalousis, L.N.; Koonen, E.; Labare, M.; Lehaut, G.; Manzanillas, L.; Mermans, J.; Michiels, I.; Moortgat, C.; Newbold, D.; Park, J.; Pestel, V.; Petridis, K.; Piñera, I.; Pommery, G.; Popescu, L.; Pronost, G.; Rademacker, J.; Ryckbosch, D.; Ryder, N.; Saunders, D.; Schune, M.-H.; Simard, L.; Vacheret, A.; Van Dyck, S.; Van Mulders, P.; van Remortel, N.; Vercaemer, S.; Verstraeten, M.; Weber, A.; Yermia, F.

2018-05-03

The SoLid collaboration has developed a new detector technology to detect electron anti-neutrinos at close proximity to the Belgian BR2 reactor at surface level. A 288 kg prototype detector was deployed in 2015 and collected data during the operational period of the reactor and during reactor shut-down. Dedicated calibration campaigns were also performed with gamma and neutron sources. This paper describes the construction of the prototype detector with a high control on its proton content and the stability of its operation over a period of several months after deployment at the BR2 reactor site. All detector cells provide sufficient light yields to achieve a target energy resolution of better than 20%/√E(MeV). The capability of the detector to track muons is exploited to equalize the light response of a large number of channels to a precision of 3% and to demonstrate the stability of the energy scale over time. Particle identification based on pulse-shape discrimination is demonstrated with calibration so...

Evaluación y estandarización del análisis por activación neutrónica según el método del k-sub cero en el reactor nuclear RP-10: Estudio preliminar empleando irradiaciones cortas

OpenAIRE

Montoya Rossi, Eduardo Haroldo

1995-01-01

Se ha estandarizado una posición de irradiación del reactor nuclear RP-10 para el uso del análisis por activación neutrónica según el método del k sub cero, empleando la convención de Högdahl y se ha evaluado el comportamiento de dicho método respecto a la exactitud y precisión de los resultados obtenidos en el análisis multielemental cuantitativo de diversos materiales certificados de referencia. Para comprobar que el método analítico se encuentra totalmente bajo control estadístico, se ha e...

Extension of the GeN-Foam neutronic solver to SP3 analysis and application to the CROCUS experimental reactor

International Nuclear Information System (INIS)

Fiorina, Carlo; Hursin, Mathieu; Pautz, Andreas

2017-01-01

Highlights: • Development and verification of an SP 3 solver based on OpenFOAM. • Integration into the GeN-Foam multi-physics platform. • Application of the new GeN-Foam SP 3 solver to the CROCUS reactor. - Abstract: The Laboratory for Reactor Physics and Systems Behaviour at the PSI and at the EPFL has been developing since 2013 a multi-physics platform for coupled reactor analysis named GeN-Foam. The developed tool includes a solver for the eigenvalue and transient solution of multi-group neutron diffusion equations. Although frequently used in reactor analysis, the diffusion theory shows some limitations for core configurations involving strong anisotropies, which is the case for the CROCUS research reactor at the EPFL. The use of an SP 3 approximation to neutron transport can often lead to visible improvements in a code predictive capabilities, especially for one-directional anisotropies, with acceptable added computational cost vs diffusion. Following some modelling issues for the CROCUS reactor, and in order to improve the GeN-Foam modelling capabilities, the GeN-Foam diffusion solver has been extended to allow for SP 3 analyses. The present paper describes such extension and a preliminary verification using a mini-core PWR benchmark. The newly developed solver is then applied to the analysis of the CROCUS experimental reactor and results are compared to Monte Carlo calculations, as well as to the results of the diffusion solver.

Full system decontamination feasibility studies

International Nuclear Information System (INIS)

Denault, R.P.; LeSurf, J.E.; Walschot, F.W.

1988-01-01

Many chemical decontaminations have been performed on subsystems in light water reactors (BWRs and PWRs) but none on the full system (including the fuel) of large, (>500 MWe) investor owned reactors. Full system decontaminations on pressure-tubed reactors have been shown to facilitate maintenance, inspection, repair and replacement of reactor components. Further advantages are increased reactor availability and plant life extension. A conceptual study has been performed for EPRI (for PWRs) and Commonwealth Edison Co (for BWRs) into the applicability and cost benefit of full system decontaminations (FSD). The joint study showed that FSDs in both PWRs and BWRs, with or without the fuel included in the decontamination, are feasible and cost beneficial provided a large amount of work is to be done following the decontamination. The large amounts of radioactive waste generated can be managed using current technologies. Considerable improvements in waste handling, and consequent cost savings, can be obtained if new techniques which are now reaching commercial application are used. (author)

Gas pollutant cleaning by a membrane reactor

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Kaldis Sotiris

2006-01-01

Full Text Available An alternative technology for the removal of gas pollutants at the integrated gasification combined cycle process for power generation is the use of a catalytic membrane reactor. In the present study, ammonia decomposition in a catalytic reactor, with a simultaneous removal of hydrogen through a ceramic membrane, was investigated. A Ni/Al2O3 catalyst was prepared by the dry and wet impregnation method and characterized by the inductively coupled plasma method, scanning electron microscopy, X-ray diffraction, and N2 adsorption before and after activation. Commercially available a-Al2O3 membranes were also characterized and the permeabilities and permselectivities of H2, N2, and CO2 were measured by the variable volume method. In parallel with the experimental analysis, the necessary mathematical models were developed to describe the operation of the catalytic membrane reactor and to compare its performance with the conventional reactor. .

Risk Management for Sodium Fast Reactors.

Energy Technology Data Exchange (ETDEWEB)

Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Groth, Katrina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-01-01

Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.

Inclusive (e,e'N), (e,e'NN), (e,e'π),... reactions in nuclei

International Nuclear Information System (INIS)

Gil, A.; Oset, E.

1997-01-01

We study the inclusive (e,e'N), (e,e'NN), (e,e'π), (e,e'πN) reactions in nuclei using a Monte Carlo simulation method to treat the multichannel problem of the final state. The input consists of reaction probabilities for the different steps evaluated using microscopical many body methods. We obtain a good agreement with experiment in some channels where there is data and make predictions for other channels which are presently under investigation in several electron laboratories. The comparison of the theoretical results with experiment for several kinematical conditions and diverse channels can serve to learn about different physical processes occurring in the reaction. The potential of this theoretical tool to make prospections for possible experiments, aiming at pinning down certain reaction probabilities, is also emphasized. (orig.)

La sociología en Colombia: vocación, disciplina, profesión e historia

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Fernando Uricoechea

2001-01-01

Full Text Available Ampliación de la presentación hecha durante La Mesa Redonda La Sociología en la Perspectiva del Desarrollo Nacional Colombiano (Mayo 11,2000. En cinco apartados: 1 Vocacion e individuo, 2 Discipilina y pr~(eJión, 3 La Disciplina y sus desafios, 4 La profesión y sus desafios, y 5 La historia; el profesor Uricoechea recoge discusiones importantes para la academia: la formación científica, la profesionalización y la Política; tres dimensiones que se entrecruzan y afectan el quehacer del sociólogo, tanto en su educación, como en su desempeño (docente o profesional. Palabras clave: Sociología en Colombia, Profesión, Positivismo, Comunidad académica

Tritium management in fusion reactors

International Nuclear Information System (INIS)

Galloway, T.R.

1978-05-01

This is a review paper covering the key environmental and safety issues and how they have been handled in the various magnetic and inertial confinement concepts and reference designs. The issues treated include: tritium accident analyses, tritium process control, occupational safety, HTO formation rate from the gas-phase, disposal of tritium contaminated wastes, and environmental impact--each covering the Joint European Tokamak (J.E.T. experiment), Tokamak Fusion Test Reactor (TFTR), Russian T-20, The Next Step (TNS) designs by Westinghouse/ORNL and General Atomic/ANL, the ANL and ORNL EPR's, the G.A. Doublet Demonstration Reactor, the Italian Fintor-D and the ORNL Demo Studies. There are also the following full scale plant reference designs: UWMAK-III, LASL's Theta Pinch Reactor Design (RTPR), Mirror Fusion Reactor (MFR), Tandem Mirror Reactor (TMR), and the Mirror Hybrid Reactor (MHR). There are four laser device breakeven experiments, SHIVA-NOVA, LLL reference designs, ORNL Laser Fusion power plant, the German ''Saturn,'' and LLL's Laser Fusion EPR I and II

FPGA based computation of average neutron flux and e-folding period for start-up range of reactors

International Nuclear Information System (INIS)

Ram, Rajit; Borkar, S.P.; Dixit, M.Y.; Das, Debashis

2013-01-01

Pulse processing instrumentation channels used for reactor applications, play a vital role to ensure nuclear safety in startup range of reactor operation and also during fuel loading and first approach to criticality. These channels are intended for continuous run time computation of equivalent reactor core neutron flux and e-folding period. This paper focuses only the computational part of these instrumentation channels which is implemented in single FPGA using 32-bit floating point arithmetic engine. The computations of average count rate, log of average count rate, log rate and reactor period are done in VHDL using digital circuit realization approach. The computation of average count rate is done using fully adaptive window size moving average method, while Taylor series expansion for logarithms is implemented in FPGA to compute log of count rate, log rate and reactor e-folding period. This paper describes the block diagrams of digital logic realization in FPGA and advantage of fully adaptive window size moving average technique over conventional fixed size moving average technique for pulse processing of reactor instrumentations. (author)

Optimized Design and Discussion on Middle and Large CANDLE Reactors

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Xiaoming Chai

2012-08-01

Full Text Available CANDLE (Constant Axial shape of Neutron flux, nuclide number densities and power shape During Life of Energy producing reactor reactors have been intensively researched in the last decades [1–6]. Research shows that this kind of reactor is highly economical, safe and efficiently saves resources, thus extending large scale fission nuclear energy utilization for thousands of years, benefitting the whole of society. For many developing countries with a large population and high energy demands, such as China and India, middle (1000 MWth and large (2000 MWth CANDLE fast reactors are obviously more suitable than small reactors [2]. In this paper, the middle and large CANDLE reactors are investigated with U-Pu and combined ThU-UPu fuel cycles, aiming to utilize the abundant thorium resources and optimize the radial power distribution. To achieve these design purposes, the present designs were utilized, simply dividing the core into two fuel regions in the radial direction. The less active fuel, such as thorium or natural uranium, was loaded in the inner core region and the fuel with low-level enrichment, e.g. 2.0% enriched uranium, was loaded in the outer core region. By this simple core configuration and fuel setting, rather than using a complicated method, we can obtain the desired middle and large CANDLE fast cores with reasonable core geometry and thermal hydraulic parameters that perform safely and economically; as is to be expected from CANDLE. To assist in understanding the CANDLE reactor’s attributes, analysis and discussion of the calculation results achieved are provided.

A full scope nuclear power plant simulator for multiple reactor types with virtual control panels

International Nuclear Information System (INIS)

Yonezawa, Hisanori; Ueda, Hiroki; Kato, Takahisa

2017-01-01

This paper summarizes a full scope nuclear power plant simulator for multiple reactor types with virtual control panels which Toshiba developed and delivered. After the Fukushima DAIICHI nuclear power plants accident, it is required that all the people who are engaged in the design, manufacturing, operation, maintenance, management and regulation for the nuclear power plant should learn the wide and deep knowledge about the nuclear power plant design including the severe accident. For this purpose, the training with a full scope simulator is one of the most suitable ways. However the existing full scope simulators which are consist of the control panels replica of the referenced plants are costly and they are hard to remodel to fit to the real plant of the latest condition. That's why Toshiba developed and delivered the new concept simulator system which covers multiple referenced plants even though they have different design like BWR and PWR. The control panels of the simulator are made by combining 69 large Liquid Crystal Display (LCD) panels with touch screen instead of a control panel replica of referenced plant. The screen size of the each panel is 42 inches and 3 displays are arranged in tandem for one unit and 23 units are connected together. Each panel displays switches, indicators, recorders and lamps with the Computer Graphics (CG) and trainees operate them with touch operations. The simulator includes a BWR and a PWR simulator model, which enable trainees to learn the wide and deep knowledge about the nuclear power plant of BWR and PWR reactor types. (author)

Habituación e institucionalización del aprovechamiento y la conservación forestal comunitaria

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Mauricio Pablo Cervantes Salas

2018-01-01

Full Text Available Este trabajo expone los mecanismos históricos de producción del campo de la conservación forestal comunitaria en un ejido ubicado en la Reserva de la Biosfera de la Mariposa Monarca, caracterizado por el buen manejo de su bosque. Mediante la teoría de los campos de Bourdieu, y los conceptos de habituación e institucionalización de Berger y Luckmann, se analizó la trayectoria histórica de este ejido en el manejo exitoso de sus bosques. Los resultados sugieren que la formación colectiva del capital cultural fue central para la producción del campo de la conservación forestal comunitaria.

Use of research and test reactors for SPD development and calibration

International Nuclear Information System (INIS)

LaFontaine, M.W.R.

2011-01-01

Prior to using a research or test reactor for performance studies or calibration of self powered detectors, it is first necessary to fully characterize the reactor environment in the region to be utilized. This presentation details Characterization Experiments performed to quantify research/test reactor core/site parameters as they would apply for use with SPD applications. Methods will be described to: Determine the Westcott parameter, r (T n /T o ) , for the region of interest; Characterize the neutron energy spectrum in terms of the cadmium absorption cut-off, i.e., consider neutrons of energy 5kT 0.13 eV to be epithermal neutrons; Determine T n , the effective neutron temperature, in the region of interest; Determine the gamma flux in the region of interest; and, Establish SPD calibration standard detectors.

Dimethyl (E-2-(N-phenylacetamidobut-2-enedioate

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Ting Bin Wen

2011-01-01

Full Text Available The title compound, C14H15NO5, was obtained from the reaction of acetanilide with dimethyl acetylenedicarboxylate in the presence of potassium carbonate. The C=C double bond adopts an E configuration and the geometry around the amide N atom is almost planar rather than pyramidal (mean deviation of 0.0032 Å from the C3N plane. The packing of the molecules in the crystal structure is stabilized by intermolecular C—H...O hydrogen bonds.

Peculiarities of approximation for reactor neutron energy spectra during computerized simulation of radiation defects

International Nuclear Information System (INIS)

Kupchishin, A.A.; Kupchishin, A.I.; Stusik, G.; Omarbekova, Zh.

2001-01-01

Peculiarities of approximation for reactor neutron energy spectra during radiation defects computerized simulation were discussed. Approximation of neutron spectra N(E) was carried out by N(E)=α·exp(-β·E)·sh(γ·E) formula (1), where α, β, γ - approximation coefficients. In the capacity of operating reactor data experimental data on 235 U and 239 Pu were applied. The algorithm was designed, and acting soft ware for spectra parameters calculation was developed. The following values of approximation parameters were obtained: α=80.8; β=0.935;γ=2.04 (for uranium and plutonium these coefficients are less distinguishing). Then with use of formula 1 and α, β, γ coefficients the approximation curves were constructed. These curves satisfactorily describe existing experimental data and allowing to use its for radiation defects simulation in the reactor materials

Neutron capture cross section measurement of $^{238}$U at the n_TOF CERN facility in the energy region from 1 eV to 700 keV

CERN Document Server

Mingrone, F; Vannini, G; Colonna, N; Gunsing, F; Zugec, P; Altstadt, S; Andrzejewski, J; Audouin, L; Barbagallo, M; Becares, V; Becvavr, F; Belloni, F; Berthoumieux, E; Billowes, J; Bosnar, D; Brugger, M; Calviani, M; Calvino, F; Cano-Ott, D; Carrapico, C; Cerutti, F; Chiaveri, E; Chin, M; Cortes, G; Cortes-Giraldo, M A; Diakaki, M; Domingo-Pardo, C; Duran, I; Dressler, R; Eleftheriadis, C; Ferrari, A; Fraval, K; Ganesan, S; Garcia, A R; Giubrone, G; Goncalves, I F; Gonzalez-Romero, E; Griesmayer, E; Guerrero, C; Hernandez-Prieto, A; Jenkins, D G; Jericha, E; Kadi, Y; Kappeler, F; Karadimos, D; Kivel, N; Koehler, P; Kokkoris, M; Krticka, M; Kroll, J; Lampoudis, C; Langer, C; Leal-Cidoncha, E; Lederer, C; Leeb, H; Leong, L S; Lo Meo, S; Losito, R; Mallick, A; Manousos, A; Marganiec, J; Martinez, T; Mastinu, P F; Mastromarco, M; Mendoza, E; Mengoni, A; Milazzo, P M; Mirea Horia, M; Mondalaers, W; Paradela, C; Pavlik, A; Perkowski, J; Plompen, A; Praena, J; Quesada, J M; Rauscher, T; Reifarth, R; Riego, A; Robles, M S; Rubbia, C; Sabate-Gilarte, M; Sarmento, R; Saxena, A; Schillebeeckx, P; Schmidt, S; Schumann, D; Tagliente, G; LTain, J; Tarrio, D; Tassan-Got, L; Tsinganis, A; Valenta, S; Variale, V; Vaz, P; Ventura, A; Vermeulen, M J; Vlachoudis, V; Vlastou, R; Wallner, A; Ware, T; Weigand, M; Weiss, C; Wright, T

2016-01-01

The aim of this work is to provide a precise and accurate measurement of the $^{238}$U(n,$\\gamma$) reaction cross section in the energy region from 1 eV to 700 keV. This reaction is of fundamental importance for the design calculations of nuclear reactors, governing the behaviour of the reactor core. In particular, fast reactors, which are experiencing a growing interest for their ability to burn radioactive waste, operate in the high energy region of the neutron spectrum. In this energy region most recent evaluations disagree due to inconsistencies in the existing measurements of up to 15%. In addition, the assessment of nuclear data uncertainty performed for innovative reactor systems shows that the uncertainty in the radiative capture cross-section of $^{238}$U should be further reduced to 1-3% in the energy region from 20 eV to 25 keV. To this purpose, addressed by the Nuclear Energy Agency as a priority nuclear data need, complementary experiments, one at the GELINA and two at the n_TOF facility, were pr...

Accident progression event tree analysis for postulated severe accidents at N Reactor

International Nuclear Information System (INIS)

Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M.; Medford, G.T.

1990-06-01

A Level II/III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The accident progression analysis documented in this report determines how core damage accidents identified in the Level I PRA progress from fuel damage to confinement response and potential releases the environment. The objectives of the study are to generate accident progression data for the Level II/III PRA source term model and to identify changes that could improve plant response under accident conditions. The scope of the analysis is comprehensive, excluding only sabotage and operator errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study. The accident progression model allows complex interactions and dependencies between systems to be explicitly considered. Latin Hypecube sampling was used to assess the phenomenological and systemic uncertainties associated with the primary and confinement system responses to the core damage accident. The results of the analysis show that the N Reactor confinement concept provides significant radiological protection for most of the accident progression pathways studied

Physicians' attitudes towards ePrescribing – evaluation of a Swedish full-scale implementation

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Montelius Emelie

2009-08-01

Full Text Available Abstract Background The penetration rate of Electronic Health Record (EHR systems in health care is increasing. However, many different EHR-systems are used with varying ePrescription designs and functionalities. The aim of the present study was to evaluate experienced ePrescribers' attitudes towards ePrescribing for suggesting improvements. Methods Physicians (n = 431 from seven out of the 21 Swedish health care regions, using one of the six most widely implemented EHR-systems with integrated electronic prescribing modules, were recruited from primary care centers and hospital clinics of internal medicine, orthopaedics and surgery. The physicians received a web survey that comprised eight questions on background data and 19 items covering attitudes towards ePrescribing. Forty-two percent (n = 199 of the physicians answered the questionnaire; 90% (n = 180 of the respondents met the inclusion criteria and were included in the final analysis. Results A majority of the respondents regarded their EHR-system easy to use in general (81%, and for the prescribing of drugs (88%. Most respondents believed they were able to provide the patients better service by ePrescribing (92%, and regarded ePrescriptions to be time saving (91% and to be safer (83%, compared to handwritten prescriptions. Some of the most frequently reported weaknesses were: not clearly displayed price of drugs (43%, complicated drug choice (21%, and the perception that it was possible to handle more than one patient at a time when ePrescribing (13%. Moreover, 62% reported a lack of receipt from the pharmacy after successful transmission of an ePrescription. Although a majority (73% of the physicians reported that they were always or often checking the ePrescription a last time before transmitting, 25% declared that they were seldom or never doing a last check. The respondents suggested a number of improvements, among others, to simplify the drug choice and the cancellation of e

Utilization of the {sup 93}Nb(n,n'){sup 93}Nb{sup m} Reaction for for Reactor Neutron Dosimetry

Energy Technology Data Exchange (ETDEWEB)

Czock, K. H.; Houtermans, H. [International Atomic Energy Agency, Division of Research and Laboratories, Vienna (Austria)

1974-09-15

For the measurement of large ({Phi}>10{sup 18}n/cm{sup 2}) fast neutron fluences in research reactors, the reaction {sup 93}Nb(n,n'){sup 93}Nb{sup m} is fairly seldom used (Lloret, Hegedus). The reason is that the excitation function for the production of the {sup 93}Nb isomeric state (half-life{approx_equal}11.4 years) by inelastic neutron scattering is nearly unknown. Also the determination of the absolute {sup 93}Nb{sup m} activity is difficult, e.g. due to the bad knowledge of the conversion coefficients {alpha}{sub tot} and {alpha}{sub K}. But for neutron dosimetry purposes one does not necessarily need absolute activity values, if one irradiated niobium foil (e.g. 10 mm empty and 0.1 mm thick) is accepted to be a reference source to other irradiated niobium foils (of the same original material). If the cross section measurements and the fluence determinations are both based on the same reference source {sup 93}Nb{sup m} as an arbitrary activity standard, the possible inaccuracy of this activity value would not matter. Such an accepted reference source could be deposited at one central institution. Once the reference source activity is defined (at present within 30% of the true absolute activity value) niobium sources with activities determined relative to this reference source could be supplied together with non-irradiated niobium foils to interested laboratories. (author)

Two-dimensional full-core transport theory Benchmarks for the WWER reactors

International Nuclear Information System (INIS)

Petkov, P.T.

2002-01-01

Several two-dimensional full-core real geometry many-group steady-state problems for the WWER-440 and WWER-1000 reactors have been solved by the MARIKO code, based on the method of characteristics. The reference transport theory solutions include assembly-wise and pin-wise power distributions. Homogenized two-group diffusion parameters and discontinuity factors have been calculated by MARIKO for each assembly type both for the whole assembly and for each cell in the smallest sector of symmetry, using the B1 method for calculation of the critical spectrum. Accurate albedo-type boundary conditions have been calculated by MARIKO for the core-reflector and core-absorber boundaries, both for each outer assembly face and for each outer cell face. Comparison with the reference solutions of the two-group nodal diffusion code SPPS-1.6 and the few-group fine-mesh diffusion codes HEX2DA and HEX2DB are presented (Authors)

Corrosión de componentes de aluminio en el reactor RP-10

OpenAIRE

Morales Larrea, Soledad; Tenorio de la Cruz, Favio

1983-01-01

Se analiza la velocidad de corrosión, por pérdida de peso de placas de aluminio a diversos valores de pH y temperatura en solución acuosa. El estudio se hace simulando las condiciones de trabajo (pH y temperatura) a las que pueden verse sometidas las vainas de aluminio de los elementos combustibles del reactor RP-10.

NARRACIÓN E IMAGINARIOS IDENTITARIOS. PARADOJAS Y PISTAS DE REFLEXIÓN

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Hermann Herlinghaus

2002-01-01

Full Text Available En la discusión actual que busca explorar arqueologías e historizar epistemologías de las nociones sujeto/subjetividad, dirigida hacia una revisión crítica de la razón etnocéntrica y especulativa, se juntan niveles y problemas diversos. Nuestra breve reflexión se va a dedicar a un concepto que por un lado ha padecido, como pocos, los lastres hegemónicos de un discurso de la modernidad, y que, por otro lado, se vincula con el campo de las prácticas sociales de una manera ambigua. Esa ambigüedad puede ser tal que, relacionada a los fenómenos de constitución de identidades, tiende a suspender la posibilidad misma de una racionalización abstracta del quehacer sociocultural. El concepto que aquí nos interesa es el de la narración.

Effect of the presence of inorganic salts on the photocatalytic inactivation of E. Coli in water Efecto de la presencia de sales inorgánicas sobre la inactivación fotocatalítica de E. Coli en agua

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Edwing Velasco

2013-03-01

Full Text Available This article presents the effect of inorganic salts MgSO4, NaCl and CaCO3 on the photocatalytic water disinfection. TiO2-P25 was used as a photocatalyst, and E. Coli was used as a contaminant. Disinfection tests were performed by controlling lighting of batch reactors loaded with contaminated water, salts and TiO2. The results of these tests were used to determine the kinetic parameters of a type Langmuir-Hinshelwood model. It was found that the salts have a strong influence on the photocatalytic inactivation of E. Coli and that each salt and its concentration affect disinfection differently in the following order: NaCl>CaCO3>>MgSO4. Additionally, the value of the calculated parameters was different for each salt, showing that the salts affect the process by several mechanisms related to the ion-bacteria interactions, ion-oxidizing species and ion-TiO2.En este artículo se presenta el efecto de las sales inorgánicas MgSO4, NaCl y CaCO3 en la desinfección fotocatalítica del agua. Se usó TiO2-P25 como fotocatalizador y E. Coli como microorganismo contaminante. Las pruebas de desinfección se realizaron mediante la iluminación controlada de reactores batch cargados con agua contaminada, sales y TiO2. Los resultados de estas pruebas fueron usados para determinar los parámetros cinéticos de un modelo tipo Langmuir-Hinshelwood. Se encontró que las sales tienen una fuerte influencia sobre la inactivación fotocatalítica de E. Coli, y que cada sal y su concentración afectan la desinfección de forma diferente y en el siguiente orden: NaCl>CaCO3>>MgSO4. Adicionalmente, el valor de los parámetros calculados fue diferente para cada sal, evidenciando que las sales afectan el proceso por varios mecanismos relacionados con las interacciones ion-bacteria, ion-especie oxidante e ion-TiO2.

La interpretación e integración de los instrumentos de Derecho Comercial

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Maximiliano Rodríguez Fernández

2013-12-01

Full Text Available El proceso de armonización a que se encuentran sometidas las normas del Derecho Comercial Internacional se realiza usualmente en tres etapas: concepción, redacción y aceptación del instrumento; adopción del instrumento internacional en los ordenamientos locales, e interpretación (aplicación del instrumento internacional. El presente artículo se refiere a la tercera, y tal vez más importante de esas etapas, esto es, la de interpretación internacional de los textos de Derecho uniforme o armónico, y particularmente lo que será la interpretación de las Reglas de Rotterdam.

Tipos de pseudobulbos e número de nós no enraizamento e brotação de Dendrobium nobile

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Ximena Maira de Souza Vilela

2011-05-01

Full Text Available Dendrobium nobile (‘Olho-de-boneca’ é uma das espécies mais cultivadas e colecionadas da família das orquidáceas. Com o objetivo de verificar a influência de reguladores de crescimento na indução de enraizamento e brotação em estacas de orquídea, testaram-se diferentes concentrações de ANA e AIB. Os pseudobulbos, também conhecidos como estacas, foram retirados de touceiras cultivadas em tronco de árvore sem nenhum tratamento prévio. No mesmo dia as folhas foram retiradas e as estacas imersas em água por 18 horas. Após este período, as estacas foram higienizadas e divididas em três partes: basal, mediana e apical, com número de gemas variando entre três e cinco. O delineamento experimental foi inteiramente casualizado, constituído de: 12 tratamentos (3 tipos de estacas x 4 concentrações de AIB, com quatro repetições com cinco estacas cada e 12 tratamentos (3 números de nós x 4 concentrações de ANA. Após a imersão por 3 minutos em solução de ANA e AIB, as estacas foram colocadas em 48 bandejas plásticas perfuradas, preenchidas com substrato de casca de arroz carbonizada e mantidas em casa de vegetação com irrigação por microaspersão. Estacas medianas, sem imersão em AIB e estacas contendo dois nós, sem imersão de ANA, proporcionam resultados mais satisfatórios na obtenção de mudas de Dendrobium nobile

The 5th surveillance testing for Kori unit 1 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

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

2000-08-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed primarily by Korea Atomic Energy Research Institute and Westinhouse corporation partially involved in testing and calculation data evaluation in order to obtain reliable test result. Fast neutron fluences for capsule V, T, S, R and P were 5.087E+18, 1.115E+19, 1.228E+19, 2.988E+19, and 3.938E+19n/cm2, respectively. The bias factor, the ratio of calculation/measurement, was 0.940 for the 1st through 5th testing and the calculational uncertainty, 7% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.9846E+19n/cm{sup 2} based on the end of 17th fuel cycle and it was predicted that the fluences of vessel inside surface at 24, 32, 40 and 48EFPY would reach 3.0593E+19, 4.0695E+19, 5.0797E+19 and 6.0900E+19n/cm{sup 2} based on the current calculation. PTS analysis for Kori unit 1 showed that 27.93EFPY was the threshold value for 300 deg F requirement. 71 refs., 33 figs., 52 tabs. (Author)

ACEPTACIÓN DEL E-COMMERCE EN COLOMBIA: UN ESTUDIO PARA LA CIUDAD DE MEDELLÍN

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JUAN F. TAVERA MESÍAS

2011-01-01

Full Text Available El e-Commerce es un fenómeno creciente en Latinoamérica y Colombia por lo que el estudio de su aceptación tecnológica es de alta importancia académica y empresarial. El objetivo de este artículo es identificar los antecedentes de la intención de uso de e-Commerce en Colombia. El Modelo deAceptación Tecnológica (TAM, es complementado en este estudio con los constructos de Confianza y Seguridad Percibida para proponer un modelo ajustado al caso colombiano. Dicho modelo es contrastado empíricamente con una muestra de consumidores de la ciudad de Medellín. Se evidencia la importancia de la confianza y la utilidad percibida como antecedentes directos de la intención de uso del e-Commerce. Los resultados llenan vacíos existentes en la literatura sobre el e-Commerce en Colombia y permiten identificar implicaciones empresariales relevantes para el desarrollo de actividades comerciales a través de la Internet.

IRPhE/STEK, Reactor Physics Experiments from Fast-Thermal Coupled Facility

International Nuclear Information System (INIS)

Dietze, Klaus; Klippel, Henk Th.; Koning, Arjan; Jacqmin, Robert

2003-01-01

1 - Description: The STEK-experiments have been performed to check neutron data of the most important reactor materials, especially of fission product nuclides, fuel isotopes and structural materials. The measured central reactivity worths (CRW) of small samples were compared with calculated values. These C/E-ratios have been used then for data corrections or in adjustment procedures. The reactors STEK (ECN Petten/ Netherlands) was a fast-thermal coupled facility of zero power. The annular thermal drivers were filled by fuel assemblies and moderated by water. The inner insertion lattices were loaded with pellets of fuel and other materials producing the fast neutron flux. The characteristics of the neutron and adjoint spectra were obtained by special arrangements of these pellets in unit cells. In this way, a hard or soft neutron spectrum or a special energy behavior of the adjoint function could be reached. The samples were moved by means of tubes to the central position (pile-oscillation technique). The original information about the facility and measurements is compiled in RCN-209, ECN-10 The 5 STEK configurations cover a broad energy range due to their increasing softness. The experiments are very valuable because of the extensive program of sample reactivity measurements with many fission product nuclides important in reactor burn-up calculations. At first, analyses of the experiments have been performed in Petten. Newer analyses were done later in Cadarache / CEA France using the European scheme for reactor calculation JEF-2.2 / ECCO / ERANOS (see Note Techniques and JEF/DOC-746). Furthermore, re-analyses were performed in O-arai / JNC Japan with the JNC standard route JENDL-3.2 / SLAROM / CITATION / PERKY. Results obtained with both code systems and different data evaluations (JEF-2.2 and JENDL-3.2) are compared in JEF/DOC-861. It contains the following documents: 31 Reports, 2 publications, 5 JEF documents, 4 conferences. 2 - Related or auxiliary programs

The timing of reactor dismantling

International Nuclear Information System (INIS)

Roberts, P.

2000-01-01

Work has been progressing across the world for the decommissioning of nuclear reactors. The initial work focused on the early, complete dismantling but this was associated with small size reactors and was done for experimental or demonstration purposes. The situation now is that an increasing number of full size power reactors are being shutdown and decision are being made as to the decommissioning strategy to be applied, e.g. with respect to the appropriate timing of reactor dismantling. There are two basic approaches to the timing of reactor dismantling, which are to either proceed with dismantling on an early time scale or to delay it for a period of years. There are a number of examples worldwide of both approaches being taken but one common feature of the approach taken by most countries is that decisions are made on a case by case basis, taking account of relevant factors, and as a result the strategy can vary from reactor to reactor and from country to country. Decisions on timing take account of the following main factors: safety, radioactive decay, financial factors, radioactive waste, reactor type, technology, repository availability, site re-use, regulatory standards, plant knowledge/records, other issues

Funciones de la vitamina E: Actualización

Directory of Open Access Journals (Sweden)

Carmen Febles Fernández

2002-04-01

Full Text Available La vitamina E agrupa diferentes compuestos, dentro de los cuales se incluyen los tocoferoles y los tocotrienoles. El más importante en la especie humana es el RRR-alfa-tocoferol. Una de las funciones más importantes atribuidas a la vitamina E es su acción antioxidante. No obstante, se han observado otras no relacionadas con esta acción. Entre estas se encuentran sus efectos sobre la proliferación celular y la acción fagocítica en el sistema inmune, que a su vez se relacionan con el efecto de esta vitamina como mensajero del estado oxidativo celular. Paralelamente, existen evidencias de su relación con la apoptosis. En este trabajo pretendemos profundizar en estas funciones de la vitamina E por sus enormes potencialidades de utilización en la terapia de enfermedades médicas y estomatológicas.Vitamin E comprises different compounds such as tocopherols and tocotrienols. The most important in the human species is RRR-alpha-tocopherol. One of the main functions attributed to vitamin E is its antioxidant action. However, other functions have been observed that are not connected with this action, as its effects on cellular proliferation and the fagocitic action in the immune system that is related at the same time to the effect of this vitamin as a messenger of the oxidative cellular state. Parallel, there are other evidences of its relationship with apoptosis. In this paper we intend to go deep into these functions of vitamin E due to its great potentialities to be used in the therapy of medical and dental diseases.

MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

International Nuclear Information System (INIS)

Ishii, M.; Revankar, S. T.; Downar, T.; Xu, Y.; Yoon, H. J.; Tinkler, D.; Rohatgi, U. S.

2003-01-01

OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral

Metallographic examination of damaged N reactor spent nuclear fuel element SFEC5,4378

Energy Technology Data Exchange (ETDEWEB)

Marschman, S.C.; Pyecha, T.D.; Abrefah, J.

1997-08-01

N-Reactor spent nuclear fuel (SNF) is currently residing underwater in the K Basins at the Hanford site, in Richland, Washington. This report presents results of the metallographic examination of specimens cut from an SNF element (Mark IV-E) with breached cladding. The element had resided in the K-West (KW) Storage Basin for at least 10 years after it was discharged from the N-Reactor. The storage containers in the KW Basin were nominally closed, isolating the SNF elements from the open pool environment. Seven specimens from this Mark IV-E outer fuel element were examined using an optical metallograph. Included were two specimens that had been subjected to a conditioning process recommended by the Independent Technical Assessment Team, two specimens that had been subjected to a conditioning process recommended in the Integrated Process Strategy Report, and three that were in the as-received, as-cut condition. One of the as-received specimens had been cut from the damaged (or breached) end of the element. All other specimens were cut from the undamaged mid-region of the fuel element. The specimens were visually examined to (1) identify uranium hydride inclusions present in the uranium metal fuel, (2) measure the thickness of the oxide layer formed on the uranium edges and assess the apparent integrity and adhesion of the oxide layer, and (3) look for features in the microstructure that might provide an insight into the various corrosion processes that occurred during underwater storage in the KW Basin. These features included, but were not limited to, the integrity of the cladding and the fuel-to-cladding bond, obvious anomalies in the microstructure, excessive pitting or friability of the fuel matrix, and obvious anomalies in the distribution of uranium hydride or uranium carbide inclusions. Also, the observed metallographic features of the conditioned specimens were compared with those of the as-received (unconditioned) specimens. 11 refs., 93 figs., 2 tabs.

Average cross-sections for /n, p/ reactions on calcium in a fission-type reactor spectrum

International Nuclear Information System (INIS)

Bruggeman, A.; Maenhaut, W.; Hoste, J.

1974-01-01

The average cross-section in a fission-type reactor spectrum sigmasub(F) was experimentally determined for the reactions 42 Ca/n,p/ 42 K, 43 Ca/n,p/ 43 K and 44 Ca/n,p/ 44 K. Calcium carbonate samples and fast neutron flux monitors were irradiated with and without cadmium shielding in the Thetis reactor (Institute for Nuclear Sciences, Rijksuniversiteit Gent). The potassium activities induced in the calcium carbonate samples were separated and purified by tetraphenylborate precipitation, after which they were measured with a Ge/Li/-detector of calibrated detection efficiency. On the basis of sigmasub(F)=0.64 mb for the reaction 27 Al/n,α/ 24 Na, the average cross-sections were as follows: 42 Ca/n,p/ 42 K: 2.82+-0.07 mb; 43 Ca/n,p/ 43 K: 1.89+-0.05 mb; 44 Ca/n,p/ 44 K: 0.065+-0.003 mb. (T.G.)

Em nós: hipertexto e literatura

Directory of Open Access Journals (Sweden)

Miguel Rettenmaier

2012-09-01

Full Text Available http://dx.doi.org/10.5007/2176-8552.2011nesp1p139 De início, gostaria de fazer uma pequena proposta, a qual, de alguma forma, motiva este trabalho: que não nos percamos, na leitura do texto literário, no dinamismo excessivo da vida atual, intimamente associada às tecnologias e aos computadores. Presumivelmente não há microtoxinas em nossa circulação, o que nos obrigaria, como Case, de Neouromancer, de Gibson, a correr contra o tempo pela vida em busca de determinado antídoto. Se o corpo é a carne, como bem sabe o protagonista do romance, e a carne tem lá suas exigências, pensemos que podemos e devemos, em dados momentos, ter o tempo a nosso favor. E isso implica a liberdade de pensar e sentir com maior acuidade certas coisas num tempo presente, mesmo sob as demandas instantâneas do tempo real. Em outras palavras: experimentar a possibilidade de existir diferentemente numa atualidade cujas relações entre informação e comunicação fundamentam-se na articulação de dois poderosos elementos intimamente associados: a conexão e a comutação.

Evaluación de los parámetros cinéticos de la ecuación de Monod

Directory of Open Access Journals (Sweden)

Alberto Duarte Torres

1996-01-01

Full Text Available La evaluación de la cinética de crecimiento de células microbiales, animales, o vegetales constituye un aspecto fundamental en el diseño, operación, simulación y predicción del comportamiento de los reactores biológicos. El propósito de este articulo es presentar una aplicación de los métodos diferencial e integral de análisis de datos en la determinación de los parámetros cinéticos del Modelo de Monod, uno de los modelos empleados con mayor frecuencia para relacionar el efecto de la concentración de sustrato sobre la velocidad especifica deformación de biomasa.

Comparison of neutron parameters between a CANDU and ACR reactors; Comparacao de parametros neutronicos entre um reator CANDU e um ACR

Energy Technology Data Exchange (ETDEWEB)

Dias, Gabriel H.P.; Silva, Clarysson A.M. da; Pereira, Claubia, E-mail: gabrielhpd@yahoo.com.br, E-mail: clarysson@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-11-01

CANDU (Canadian Deuterium Uranium) is a type of reactor that uses heavy water (D{sub 2}O) as a moderator and as a refrigerant. Despite having chemical characteristics similar to light water (H{sub 2}O), heavy water has a high moderation ratio for neutrons. This feature enables CANDU to use natural uranium as fuel. However, research has evaluated the possibility of using H{sub 2}O as a refrigerant and D{sub 2}O as a moderator aiming at reducing the volume of heavy water. Such changes would imply the use of lightly enriched uranium due to the presence of H{sub 2}O. In this context, the concept of ACR (Advanced CANDU Reactor) has been developed. This reactor has an innovative design which combines of the current CANDU with the characteristics of PWR (Pressurized Water Reactor) type reactors. Studies by AECL (Atomic Energy Canada Limited) show that compared to CANDU, the ACR presents a cost reduction in construction, improved firing performance, improved operation safety and longer life. The present work aims to evaluate, in steady state, some of the main neutron parameters of CANDU-6 and ACR-1000. The MCNPX 2.6.0 code (Monte Carlo N-Particle eXtended -version 2.6.0) was used to simulate such types of reactors. The results show that the models configured in the MCNPX adequately reproduce the neutron behavior of the studied reactors. These models may be used in future work for analysis of fuel burn and evolution.

EVALUACIÓN DE LA EFICIENCIA DE REACTORES DE LECHO FIJO UTILIZANDO AGUAS MIELES RESIDUALES DE TRAPICHES ARTESANALES

Directory of Open Access Journals (Sweden)

Gloria Lucía Cárdenas Calvachi

2009-01-01

Full Text Available El problema de tratamiento y disposición final de las aguas mieles residuales provenientes de los trapiches artesanales en el departamento de Nariño, en particular los asentados en el municipio de Sandoná, radica en su imposibilidad económica y tecnológica, dadas las características de subsistencia en que se basa su funcionamiento. El sistema de fi ltros anaerobios de fl ujo ascendente (FAFA como unidad principal de tratamiento biológico en la degradación de azúcares, ofrece una buena alternativa por ser considerado efi ciente, de relativos bajos costos de construcción, operación y mantenimiento, con el reto central de mantener las condiciones de hábitat adecuadas para el crecimiento de la biomasa al interior del reactor. Se evaluaron cuatro medios de contacto (concha marina, material sintético, material vitrifi cado y grava de río, a escala de laboratorio, para encontrar el lecho de soporte de FAFA más conveniente en condiciones controladas de temperatura, régimen de flujo y acondicionamiento previo del medio bacteriano. La concha marina y el material sintético, ofrecen características de resistencia, durabilidad y facilidad de consecución y alcanzan remociones de materia orgánica mayores del 80%. Sin embargo, la concha marina alcanza las mayores remociones (89,7% para DQO y 87,8 % para DBO gracias a su estructura física que ofrece un microambiente adecuado y por su composición química, fuente natural de alcalinidad y micronutrientes al sistema, lo que hace que se lo considere como el medio de contacto más adecuado para diseñar e implementar fi ltros anaerobios de lecho fi jo en la industria artesanal panelera.

The ICRH tokamak fusion test reactor

International Nuclear Information System (INIS)

Perkins, F.W.

1976-01-01

A Tokamak Fusion Test Reactor where the ion are maintained at Tsub(i) approximately 20keV>Tsub(e) approximately 7keV by ion-cyclotron resonance heating is shown to produce an energy amplification of Q>2 provided the principal ion energy loss channel is via collisional transfer to the electrons. Such a reactor produces 19MW of fusion power to the electrons. Such a reactor produces 19MW of fusion power and requires a 50MHz radio-frequency generator capable of 50MW peak power; it is otherwise compatible with the conceptual design for the Princeton TFTR. The required n tausub(E) values for electrons and ions are respectively ntausub(Ee)>1.5.10 13 cm -3 -sec and ntausub(Ei)>4.10 13 cm -3 -sec. The principal areas where research is needed to establish this concept are: tokamak transport calculations, ICRH physics, trapped-particle instability energy losses, tokamak equilibria with high values of βsub(theta), and, of course, impurities

Search for eV sterile neutrinos at a nuclear reactor — the Stereo project

Science.gov (United States)

Haser, J.; Stereo Collaboration

2016-05-01

The re-analyses of the reference spectra of reactor antineutrinos together with a revised neutrino interaction cross section enlarged the absolute normalization of the predicted neutrino flux. The tension between previous reactor measurements and the new prediction is significant at 2.7 σ and is known as “reactor antineutrino anomaly”. In combination with other anomalies encountered in neutrino oscillation measurements, this observation revived speculations about the existence of a sterile neutrino in the eV mass range. Mixing of this light sterile neutrino with the active flavours would lead to a modification of the detected antineutrino flux. An oscillation pattern in energy and space could be resolved by a detector at a distance of few meters from a reactor core: the neutrino detector of the Stereo project will be located at about 10 m distance from the ILL research reactor in Grenoble, France. Lengthwise separated in six target cells filled with 2 m3 Gd-loaded liquid scintillator in total, the experiment will search for a position-dependent distortion in the energy spectrum.

Modelo Teórico Integrado de Gamificación en Ambientes E-Learning (E-MIGA

Directory of Open Access Journals (Sweden)

Angel Torres-Toukoumidis

2018-01-01

Full Text Available La presencia de elementos lúdicos en el contexto educativo, específicamente en la modalidad de formación E-learning, demuestra que la incorporación de la gamificación sistematiza la experiencia del usuario en base a parámetros relacionados con los juegos. Este artículo presenta la integración de dos modelos conceptuales de gamificación extraídos de la revisión literaria publicada entre el 2012 y el 2015, validando cada una de las dimensiones e indicadores mediante un estudio Delphi con expertos en pedagogía y diseño de juegos. Posteriormente, se exhiben los resultados cualitativos de la aplicación del modelo en 6 aplicaciones móviles educativas analizadas durante 7 meses (junio 2015- enero 2016 manifestando la correlación de los modelos seleccionados según los criterios de idoneidad y pertinencia. En definitiva, el siguiente modelo construye una vía de conexión entre la base teórica y el análisis empírico de la gamificación con el fin de sobreponerse a los retos de la educación E-learning en el siglo XXI teniendo como última instancia la transposición del modelo teórico a un modelo de utilidad cuantitativo al servicio de futuras investigaciones.

Transportation risk assessment of radioactive wastes generated by the N-Reactor stabilization program at the Hanford Site, Washington

International Nuclear Information System (INIS)

Wheeler, T.

1994-12-01

The potential radiological and nonradiological risks associated with specific radioactive waste shipping campaigns at the Hanford Site are estimated. The shipping campaigns analyzed are associated with the transportation of wastes from the N-Reactor site at the 200-W Area, both within the Hanford Reservation, for disposal. The analysis is based on waste that would be generated from the N-Reactor stabilization program

Assessing the degree of plug flow in oxidation flow reactors (OFRs: a study on a potential aerosol mass (PAM reactor

Directory of Open Access Journals (Sweden)

D. Mitroo

2018-03-01

Full Text Available Oxidation flow reactors (OFRs have been developed to achieve high degrees of oxidant exposures over relatively short space times (defined as the ratio of reactor volume to the volumetric flow rate. While, due to their increased use, attention has been paid to their ability to replicate realistic tropospheric reactions by modeling the chemistry inside the reactor, there is a desire to customize flow patterns. This work demonstrates the importance of decoupling tracer signal of the reactor from that of the tubing when experimentally obtaining these flow patterns. We modeled the residence time distributions (RTDs inside the Washington University Potential Aerosol Mass (WU-PAM reactor, an OFR, for a simple set of configurations by applying the tank-in-series (TIS model, a one-parameter model, to a deconvolution algorithm. The value of the parameter, N, is close to unity for every case except one having the highest space time. Combined, the results suggest that volumetric flow rate affects mixing patterns more than use of our internals. We selected results from the simplest case, at 78 s space time with one inlet and one outlet, absent of baffles and spargers, and compared the experimental F curve to that of a computational fluid dynamics (CFD simulation. The F curves, which represent the cumulative time spent in the reactor by flowing material, match reasonably well. We value that the use of a small aspect ratio reactor such as the WU-PAM reduces wall interactions; however sudden apertures introduce disturbances in the flow, and suggest applying the methodology of tracer testing described in this work to investigate RTDs in OFRs to observe the effect of modified inlets, outlets and use of internals prior to application (e.g., field deployment vs. laboratory study.

E/Z-Photoisomerization of N,N'-Bis(4-dimethylaminobenzylidene1,2-diaminoethane and N,N'-Bis(4-dimethylaminobenzylidene1,3-diaminopropane in Chloroform

Directory of Open Access Journals (Sweden)

Asghar Samimi

2011-01-01

Full Text Available The E/Z-Photoisomerization of Schiff bases N,N'-bis(4-dimethylamino benzylidene1,2-diaminoethane (BDAE and N,N'-bis(4-dimethylaminobenzyli-dene1,3-diaminopropane (BDAP were studied by UV-Vis absorption spectroscopy and theoretical chemistry calculations. Photochemical investigations in solution depict the time resolved spectral changes, recorded before and after irradiation. The shift indicates the transformation from E to Z configuration of the C=N bond in solution for BDAE and BDAP. Spectra profiles and kinetic constants were evaluated using multivariate curve resolution and non-linear least squares curve fitting by toolbox of MATLAB program using the corresponding absorption spectra-time data. The experimental results show that BDAP can perform the photochromism easier than BDAE, may be due to the molecular topology difference.

15 N utilization in nitride nuclear fuels for advanced nuclear power reactors and accelerator - driven systems

International Nuclear Information System (INIS)

Axente, D.

2005-01-01

15 N utilization for nitride nuclear fuels production for nuclear power reactors and accelerator - driven systems is presented. Nitride nuclear fuel is the obvious choice for advanced nuclear reactors and ADS because of its favorable properties: a high melting point, excellent thermal conductivity, high fissile density, lower fission gas release and good radiation tolerance. The application of nitride fuels in nuclear reactors and ADS requires use of 15 N enriched nitrogen to suppress 14 C production due to (n,p) reaction on 14 N. Accelerator - driven system is a recent development merging of accelerator and fission reactor technologies to generate electricity and transmute long - lived radioactive wastes as minor actinides: Np, Am, Cm. A high-energy proton beam hitting a heavy metal target produces neutrons by spallation. The neutrons cause fission in the fuel, but unlike in conventional reactors, the fuel is sub-critical and fission ceases when the accelerator is turned off. Nitride fuel is a promising candidate for transmutation in ADS of minor actinides, which are converted into nitrides with 15 N for that purpose. Tacking into account that the world wide market is about 20 to 40 Kg 15 N annually, the supply of that isotope for nitride fuel production for nuclear power reactors and ADS would therefore demand an increase in production capacity by a factor of 1000. For an industrial plant producing 100 t/y 15 N, using present technology of isotopic exchange in NITROX system, the first separation stage of the cascade would be fed with 10M HNO 3 solution of 600 mc/h flow - rate. If conversion of HNO 3 into NO, NO 2 , at the enriching end of the columns, would be done with gaseous SO 2 , for a production plant of 100 t/y 15 N a consumption of 4 million t SO 2 /y and a production of 70 % H 2 SO 4 waste solution of 4.5 million mc/y are estimated. The reconversion of H 2 SO 4 into SO 2 in order to recycle of SO 2 is a problem to be solved to compensate the cost of SO 2

An automatic device for sample insertion and extraction to/from reactor irradiation facilities

International Nuclear Information System (INIS)

Alloni, L.; Venturelli, A.; Meloni, S.

1990-01-01

At the previous European Triga Users Conference in Vienna,a paper was given describing a new handling tool for irradiated samples at the L.E.N.A plant. This tool was the first part of an automatic device for the management of samples to be irradiated in the TRIGA MARK ii reactor and successively extracted and stored. So far sample insertion and extraction to/from irradiation facilities available on reactor top (central thimble,rotatory specimen rack and channel f),has been carried out manually by reactor and health-physics operators using the ''traditional'' fishing pole provided by General Atomic, thus exposing reactor personnel to ''unjustified'' radiation doses. The present paper describes the design and the operation of a new device, a ''robot''type machine,which, remotely operated, takes care of sample insertion into the different irradiation facilities,sample extraction after irradiation and connection to the storage pits already described. The extraction of irradiated sample does not require the presence of reactor personnel on the reactor top and,therefore,radiation doses are strongly reduced. All work from design to construction has been carried out by the personnel of the electronic group of the L.E.N.A plant. (orig.)

Evaluation of In-Core Fuel Management for the Transition Cores of RSG-GAS Reactor to Full-Silicide Core

International Nuclear Information System (INIS)

S, Tukiran; MS, Tagor; P, Surian

2003-01-01

The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 gU/cc has been done. The core-of RSG-GAS reactor has been operated full core of silicide fuels which is started with the mixed core of oxide-silicide start from core 36. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 9 transition cores (core 36 - 44) to achieve a full-silicide core (core 45). The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters. Conversion core was achieved by transition cores mixed oxide-silicide fuels. Each transition core is calculated and measured core parameter such as, excess reactivity and shutdown margin. Calculation done by Batan-EQUIL-2D code and measurement of the core parameters was carried out using the method of compensation of couple control rods. The results of calculation and experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safely to a full-silicide core

SCINFUL: A Monte Carlo based computer program to determine a scintillator full energy response to neutron detection for E/sub n/ between 0. 1 and 80 MeV: Program development and comparisons of program predictions with experimental data

Energy Technology Data Exchange (ETDEWEB)

Dickens, J.K.

1988-04-01

This document provides a discussion of the development of the FORTRAN Monte Carlo program SCINFUL (for scintillator full response), a program designed to provide a calculated full response anticipated for either an NE-213 (liquid) scintillator or an NE-110 (solid) scintillator. The program may also be used to compute angle-integrated spectra of charged particles (p, d, t, /sup 3/He, and ..cap alpha..) following neutron interactions with /sup 12/C. Extensive comparisons with a variety of experimental data are given. There is generally overall good agreement (<10% differences) of results from SCINFUL calculations with measured detector responses, i.e., N(E/sub r/) vs E/sub r/ where E/sub r/ is the response pulse height, reproduce measured detector responses with an accuracy which, at least partly, depends upon how well the experimental configuration is known. For E/sub n/ < 16 MeV and for E/sub r/ > 15% of the maximum pulse height response, calculated spectra are within +-5% of experiment on the average. For E/sub n/ up to 50 MeV similar good agreement is obtained with experiment for E/sub r/ > 30% of maximum response. For E/sub n/ up to 75 MeV the calculated shape of the response agrees with measurements, but the calculations underpredicts the measured response by up to 30%. 65 refs., 64 figs., 3 tabs.

Nuclear research reactors

International Nuclear Information System (INIS)

1985-01-01

It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

Biological biogas upgrading capacity of a hydrogenotrophic community in a trickle-bed reactor

International Nuclear Information System (INIS)

Rachbauer, Lydia; Voitl, Gregor; Bochmann, Günther; Fuchs, Werner

2016-01-01

Highlights: • Data on long term operation of a system supplied with real biogas are presented. • Ex-situ biological methanation is feasible for biogas upgrading. • Gas quality obtained complies with strictest direct grid injection criteria. • Biomethane can act as flexible storage for renewable surplus electricity. - Abstract: The current study reports on biological biogas upgrading by means of hydrogen addition to obtain biomethane. A mesophilic (37 °C) 0.058 m"3 trickle-bed reactor with an immobilized hydrogenotrophic enrichment culture was operated for a period of 8 months using a substrate mix of molecular hydrogen (H_2) and biogas (36–42% CO_2). Complete CO_2 conversion (> 96%) was achieved up to a H_2 loading rate of 6.5 m_n"3 H_2/m"3_r_e_a_c_t_o_r _v_o_l_. × d, corresponding to 2.3 h gas retention time. The optimum H_2/CO_2 ratio was determined to be between 3.67 and 4.15. CH_4 concentrations above 96% were achieved with less than 0.1% residual H_2. This gas quality complies even with tightest standards for grid injection without the need for additional CO_2 removal. If less rigid standards must be fulfilled H_2 loading rates can be almost doubled (10.95 versus 6.5 m_n"3 H_2/m"3_r_e_a_c_t_o_r _v_o_l_. × d) making the process even more attractive. At this H_2 loading the achieved methane productivity was 2.52 m_n"3 CH_4/m"3_r_e_a_c_t_o_r _v_o_l_. × d. In terms of biogas this corresponds to an upgrading capacity of 6.9 m_n"3 biogas/m"3_r_e_a_c_t_o_r _v_o_l_. × d. The conducted experiments demonstrate that biological methanation in an external reactor is well feasible for biogas upgrading under the prerequisite that an adequate H_2 source is available.

Trastorno por Déficit de Atención e Hiperactividad (TDAH: Relación con las redes atencionales

Directory of Open Access Journals (Sweden)

Juan Lupiáñez

2011-02-01

Full Text Available El trastorno por déficit de atención e hiperactividad (TDAH afecta a entre un 5% y un 10% de la población en desarrollo. Hiperactividad, impulsividad e inatención son los síntomas comportamentales que mejor caracterizan a este trastorno. A pesar de la extensa investigación clínica en relación al mismo, el déficit subyacente sigue aún generando bastante controversia. Por ello, estudiar las redes atencionales en relación a la caracterización sintomática del TDAH puede ayudar a tener una visión más clara del trastorno y con ello generar mejoras en su tratamiento.

The rehabilitation/upgrading of Philippine Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Renato, T Banaga [Philippines Nuclear Research Inst., Quezon (Philippines)

1998-10-01

The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E{sub 1}-U-Z{sub 1}-H{sub 1.6} TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

The rehabilitation/upgrading of Philippine Research Reactor

International Nuclear Information System (INIS)

Renato T, Banaga

1998-01-01

The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E 1 -U-Z 1 -H 1.6 TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Relato autobiográfico e interpretación: una concepción narrativa de la identidad personal

Directory of Open Access Journals (Sweden)

Dante G. Duero

2006-05-01

Full Text Available En este trabajo analizo el lugar de la explicación y la interpretación dentro de las ciencias en general y de las ciencias sociales e históricas en particular. Expongo los argumentos de algunos autores en favor de la especificidad y la legitimidad del modelo narrativo dentro de las ciencias sociales e históricas. A continuación analizo la tesis que propone pensar a la identidad personal como parte de un relato que presenta grados diferentes de coherencia e integración. A partir de propuestas teóricas de diferentes autores, hago una caracterización de esta clase de narraciones y los elementos que la componen. Finalmente, reviso la tesis que afirma que la psicoterapia es un espacio para la reconstrucción y reelaboración de los relatos de los pacientes, con vistas a lograr narraciones más coherentes.

Aspectos claves relacionados con la gestión del e.Learning

Directory of Open Access Journals (Sweden)

Francisco Lupiáñez Villanueva

Full Text Available La introducción y uso de las Tecnologías de la Información y la Comunicación, especialmente Internet, en el ámbito de la salud, la medicina y los sistemas sanitarios enfrenta a todos los agentes e instituciones ante numerosos retos. La educación y la formación, como actividades claves tanto de la sociedad de la información como de los sistemas sanitarios, también se ven afectadas por este proceso de cambio. La finalidad de este artículo es identificar y analizar algunos de los aspectos claves relacionados con la introducción de las TIC e Internet en el ámbito de la educación médica continuada desde el punto de vista organizacional. Esta visión nos permite combinar factores pedagógicos, tecnológicos y organizativos. Todos ellos necesarios a la hora de abordar las complejas interacciones que se producen en la implantación de un proyecto de e.Learning.

Aspectos claves relacionados con la gestión del e.Learning

Directory of Open Access Journals (Sweden)

Francisco Lupiáñez Villanueva

2006-12-01

Full Text Available La introducción y uso de las Tecnologías de la Información y la Comunicación, especialmente Internet, en el ámbito de la salud, la medicina y los sistemas sanitarios enfrenta a todos los agentes e instituciones ante numerosos retos. La educación y la formación, como actividades claves tanto de la sociedad de la información como de los sistemas sanitarios, también se ven afectadas por este proceso de cambio. La finalidad de este artículo es identificar y analizar algunos de los aspectos claves relacionados con la introducción de las TIC e Internet en el ámbito de la educación médica continuada desde el punto de vista organizacional. Esta visión nos permite combinar factores pedagógicos, tecnológicos y organizativos. Todos ellos necesarios a la hora de abordar las complejas interacciones que se producen en la implantación de un proyecto de e.Learning.

Experimental study of exclusive $^2$H$(e,e^\\prime p)n$ reaction mechanisms at high $Q^2$

Energy Technology Data Exchange (ETDEWEB)

Kim Egiyan; Gegham Asryan; Nerses Gevorgyan; Keith Griffioen; Jean Laget; Sebastian Kuhn; Gary Adams; Moscov Amaryan; Pawel Ambrozewicz; Marco Anghinolfi; Gerard Audit; Harutyun AVAKIAN; Harutyun Avakian; Hovhannes Baghdasaryan; Nathan Baillie; Jacques Ball; Nathan Baltzell; Steve Barrow; Vitaly Baturin; Marco Battaglieri; Ivan Bedlinski; Ivan Bedlinskiy; Mehmet Bektasoglu; Matthew Bellis; Nawal Benmouna; Barry Berman; Angela Biselli; Lukasz Blaszczyk; Sylvain Bouchigny; Sergey Boyarinov; Robert Bradford; Derek Branford; William Briscoe; William Brooks; Stephen Bueltmann; Volker Burkert; Cornel Butuceanu; John Calarco; Sharon Careccia; Daniel Carman; Antoine Cazes; Shifeng Chen; Philip Cole; Patrick Collins; Philip Coltharp; Dieter Cords; Pietro Corvisiero; Donald Crabb; Volker Crede; John Cummings; Natalya Dashyan; Rita De Masi; Raffaella De Vita; Enzo De Sanctis; Pavel Degtiarenko; Haluk Denizli; Lawrence Dennis; Alexandre Deur; Kahanawita Dharmawardane; Richard Dickson; Chaden Djalali; Gail Dodge; Joseph Donnelly; David Doughty; Michael Dugger; Steven Dytman; Oleksandr Dzyubak; Hovanes Egiyan; Lamiaa Elfassi; Latifa Elouadrhiri; Paul Eugenio; Renee Fatemi; Gleb Fedotov; Gerald Feldman; Robert Feuerbach; Robert Fersch; Michel Garcon; Gagik Gavalian; Gerard Gilfoyle; Kevin Giovanetti; Francois-Xavier Girod; John Goetz; Atilla Gonenc; Christopher Gordon; Ralf Gothe; Michel Guidal; Matthieu Guillo; Hayko Guler; Lei Guo; Vardan Gyurjyan; Cynthia Hadjidakis; Kawtar Hafidi; Hayk Hakobyan; Rafael Hakobyan; Charles Hanretty; John Hardie; F. Hersman; Kenneth Hicks; Ishaq Hleiqawi; Maurik Holtrop; Charles Hyde-Wright; Yordanka Ilieva; David Ireland; Boris Ishkhanov; Eugeny Isupov; Mark Ito; David Jenkins; Hyon-Suk Jo; Kyungseon Joo; Henry Juengst; Narbe Kalantarians; James Kellie; Mahbubul Khandaker; Wooyoung Kim; Andreas Klein; Franz Klein; Alexei Klimenko; Mikhail Kossov; Zebulun Krahn; Laird Kramer; V. Kubarovsky; Joachim Kuhn; Sergey Kuleshov; Jeff Lachniet; Jorn Langheinrich; David Lawrence; Ji Li; Kenneth Livingston; Haiyun Lu; Marion MacCormick; Claude Marchand; Nikolai Markov; Paul Mattione; Simeon McAleer; Bryan McKinnon; John McNabb; Bernhard Mecking; Surik Mehrabyan; Joseph Melone; Mac Mestayer; Curtis Meyer; Tsutomu Mibe; Konstantin Mikhaylov; Ralph Minehart; Marco Mirazita; Rory Miskimen; Viktor Mokeev; Kei Moriya; Steven Morrow; Maryam Moteabbed; James Mueller; Edwin Munevar Espitia; Gordon Mutchler; Pawel Nadel-Turonski; Rakhsha Nasseripour; Silvia Niccolai; Gabriel Niculescu; Maria-Ioana Niculescu; Bogdan Niczyporuk; Megh Niroula; Rustam Niyazov; Mina Nozar; Grant O' Rielly; Mikhail Osipenko; Alexander Ostrovidov; Kijun Park; Evgueni Pasyuk; Craig Paterson; Sergio Pereira; Joshua Pierce; Nikolay Pivnyuk; Dinko Pocanic; Oleg Pogorelko; Sergey Pozdnyakov; Barry Preedom; John Price; Yelena Prok; Dan Protopopescu; Brian Raue; Gregory Riccardi; Giovanni Ricco; Marco Ripani; Barry Ritchie; Federico Ronchetti; Guenther Rosner; Patrizia Rossi; Franck Sabatie; Julian Salamanca; Carlos Salgado; Joseph Santoro; Vladimir Sapunenko; Reinhard Schumacher; Vladimir Serov; Youri Sharabian; Nikolay Shvedunov; Alexander Skabelin; Elton Smith; Lee Smith; Daniel Sober; Daria Sokhan; Aleksey Stavinskiy; Samuel Stepanyan; Stepan Stepanyan; Burnham Stokes; Paul Stoler; Steffen Strauch; Mauro Taiuti; David Tedeschi; Ulrike Thoma; Avtandil Tkabladze; Svyatoslav Tkachenko; Luminita Todor; Clarisse Tur; Maurizio Ungaro; Michael Vineyard; Alexander Vlassov; Daniel Watts; Lawrence Weinstein; Dennis Weygand; M. Williams; Elliott Wolin; Michael Wood; Amrit Yegneswaran; Lorenzo Zana; Jixie Zhang; Bo Zhao; Zhiwen Zhao

2007-06-01

The reaction {sup 2}H(e,e{prime} p)n has been studied with full kinematic coverage for photon virtuality 1.75 < 5.5 {approx} GeV{sup 2}. Comparisons of experimental data with theory indicate that for very low values of neutron recoil momentum (p{sub n} < 100 MeV/c) the neutron is primarily a spectator and the reaction can be described by the plane-wave impulse approximation. For 100 < 750 MeV/c proton-neutron rescattering dominates the cross section, while {Delta} production followed by the N{Delta} {yields} NN transition is the primary contribution at higher momenta.

El sistema de evaluación de impacto ambiental y la implementación de los instrumentos de regulación territorial urbanos e interurbanos

Directory of Open Access Journals (Sweden)

Pedro Lira Olmo

1997-08-01

Full Text Available A continuación se presentan algunas consideraciones en relación a lo establecido por la Ley de Bases del Medio Ambiente, para la evaluación ambiental de los instrumentos de regulación territorial urbana e interurbana. La inquietud que motiva esta ponencia radica en que publicado el Reglamento del Sistema de Evaluación de Impacto Ambiental con fecha 3 de abril de 1997, el Sistema se vuelve obligatorio e incluye a los instrumentos en comento. Esta situación presenta, como se expondrá, singularidades que será preciso definir por la autoridad, para permitir un desarrollo fluido del SEIA y los Planes de Desarrollo Urbano e Interurbano.

N2O Catalytic Decomposition – from Laboratory Experiment to Industry Reactor

Czech Academy of Sciences Publication Activity Database

Obalová, L.; Jirátová, Květa; Karásková, K.; Chromčáková, Ž.

2012-01-01

Roč. 191, č. 1 (2012), s. 116-120 ISSN 0920-5861 R&D Projects: GA TA ČR TA01020336 Institutional support: RVO:67985858 Keywords : N2O * catalytic decomposition * fixed bed reactor Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 2.980, year: 2012

Three-dimensional reactor model for the Paks NPP full-scope simulator

International Nuclear Information System (INIS)

Gyori, C.; Hegyi, G.; Hozer, Z.; Kereszturi, A.; Maraczy, C.

1993-01-01

The reactor model includes thermohydraulic and neutron-physical components. The thermohydraulic model is based on the SMABRE code developed at the Technical Research Centre of Finland for the analysis of loss-of-coolant transients in PWRs. The fuel rod model will be replaced by a new software module providing a comprehensive description of the behavior of fuel rods during reactor transients and hypothetical accidents. The calculation is performed in four individual models: fuel rod temperature model, fuel rod internal pressure model, fuel rod deformation model and fuel rod failure model. In the neutron-physical model the core is calculated with nodes for all of the 349 fuel assemblies, and each assembly is calculated in ten layers. (Z.S.) 1 fig., 5 refs

Estudio de la distribución de los tiempos de residencia en un reactor tubular para la hidrólisis de lecitina de soja con fosfolipasa A2 inmovilizada

Directory of Open Access Journals (Sweden)

Zaritzky, N.

2001-10-01

Full Text Available The hydrolisis of soybean lecithin can be carried out by means of the use of immobilized A2 phospholipase which releases a fatty acid of C-2 position of the phospholipids so that an enriched product in lysolecithins is obtained. The enzymatic reaction follows a first order kinetics when the substrate concentrations are in the range: 6.34•10-3 M to 19.0•10-3 M. The value of the rate constant: k= 9.88•10-2 min-1 corresponds to the one obtained for the immobilized enzyme on alumina. A reactor was constructed and alumina was the selected support because of its good mechanic properties and fundamentally because of its low cost. The flow behaviour in the reactor and how it departs from the ideal model of plug-flow was analyzed by injecting a NaCl solution of a well-known concentration (tracer and then, passing it through the reactor. According to the experiences carried out, the conductivity measures proved adequate for the determination of the residence times. The system showed lineal behaviour. The residence times in the experimentally built reactor for different load arrangements (particle support + inert load was analyzed by using three different flows. The nonconverted fractions for the reactor were calculated and differences in the output were observed, in comparison to the plug-flow reactor, precisely because of channelizations and shut-offs that are generated inside the column. Maximal conversion in the experiences carried out both with higher substrate concentrations and for a minor feed flow were achieved. The dispersion module resulted quite higher than the limit that introduces a gaussian curve, for the one for which the degree of supposition of high dispersion was correct. The reactor showed a behaviour similar to that of a reactor of complete mixture and it was concluded that the degree of back-mixing, the formation of whirls and zones of redistribution of material are important.La hidrólisis de lecitina de soja puede ser

Hanford Site 100-N Area In Situ Bioremediation of UPR-100-N-17, Deep Petroleum Unplanned Release - 13245

International Nuclear Information System (INIS)

Saueressig, Daniel G.

2013-01-01

In 1965 and 1966, approximately 303 m 3 of Number 2 diesel fuel leaked from a pipeline used to support reactor operations at the Hanford Site's N Reactor. N Reactor was Hanford's longest operating reactor and served as the world's first dual purpose reactor for military and power production needs. The Interim Action Record of Decision for the 100-N Area identified in situ bioremediation as the preferred alternative to remediate the deep vadose zone contaminated by this release. A pilot project supplied oxygen into the vadose zone to stimulate microbial activity in the soil. The project monitored respiration rates as an indicator of active biodegradation. Based on pilot study results, a full-scale system is being constructed and installed to remediate the vadose zone contamination. (authors)

A WIMS E analysis of zero energy experiments performed on the Dragon reactor

Energy Technology Data Exchange (ETDEWEB)

Lancefield, M. J.; Broadhouse, B.; Woloch, F.

1974-10-15

UKAEA methods embodied in the WINS-E modular scheme of codes are described in their application to the analysis of zero energy experiments performed on the DRAGON reactor. Measured reactivity and reaction rate distributions are compared with the predictions of the analysis.

Power measurement of the RA-3 reactor using the neutron noise technique and 16N

International Nuclear Information System (INIS)

Gomez, Angel

2003-01-01

This work describes a measurement method based on the neutron noise technique which is used for determining the relation between the power and the currents of two ionization chambers. These chambers are sensitive to the gamma radiation from the 16 N decay produced in the RA-3 reactor core. The power during operation is obtained from the calibration factors by measuring those currents. As this calibration factors depend on the cooler flow that circulates in the reactor core and in the 16 N measuring system, an estimator, that is a function of the ratio of this currents, is proposed in order to detect flow changes. (author)

Serious accidents of PWR type reactors for power generation

International Nuclear Information System (INIS)

2008-12-01

This document presents the great lines of current knowledge on serious accidents relative to PWR type reactors. First, is exposed the physics of PWR type reactor core meltdown and the possible failure modes of the containment building in such a case. Then, are presented the dispositions implemented with regards to such accidents in France, particularly the pragmatic approach that prevails for the already built reactors. Then, the document tackles the case of the European pressurized reactor (E.P.R.), for which the dimensioning takes into account explicitly serious accidents: it is a question of objectives conception and their respect must be the object of a strict demonstration, by taking into account uncertainties. (N.C.)

Environmental aspects of fusion reactors

International Nuclear Information System (INIS)

Coffman, F.E.; Williams, J.M.

1975-01-01

With the continued depletion of fossil and uranium resources in the coming decades, the U. S. will be forced to look more toward renewable energy resources (e.g., wind, tidal, geothermal, and solar power) and toward such longer-term and nondepletable energy resources as fissile fast breeder reactors and fusion power. Several reference reactor designs have been completed for full-scale fusion power reactors that indicate that the environmental impacts from construction, operation, and eventual decommissioning of fusion reactors will be quite small. The principal environmental impact from fusion reactor operation will be from thermal discharges. Some of the safety and environmental characteristics that make fusion reactors appear attractive include an effectively infinite fuel supply at low cost, inherent incapability for a ''nuclear explosion'' or a ''nuclear runaway,'' the absence of fission products, the flexibility of selecting low neutron-cross-section structural materials so that emergency core cooling for a loss-of-coolant or other accident will not be necesary, and the absence of special nuclear materials such as 235 U or 239 Pu, so that diversion of nuclear weapons materials will not be possible and nuclear blackmail will not be a serious concern

Full scale seismic simulation of a nuclear reactor with parallel finite element analysis code for assembled structure

International Nuclear Information System (INIS)

Yamada, Tomonori

2010-01-01

The safety requirement of nuclear power plant attracts much attention nowadays. With the growing computing power, numerical simulation is one of key technologies to meet this safety requirement. Center for Computational Science and e-Systems of Japan Atomic Energy Agency has been developing a finite element analysis code for assembled structure to accurately evaluate the structural integrity of nuclear power plant in its entirety under seismic events. Because nuclear power plant is very huge assembled structure with tens of millions of mechanical components, the finite element model of each component is assembled into one structure and non-conforming meshes of mechanical components are bonded together inside the code. The main technique to bond these mechanical components is triple sparse matrix multiplication with multiple point constrains and global stiffness matrix. In our code, this procedure is conducted in a component by component manner, so that the working memory size and computing time for this multiplication are available on the current computing environment. As an illustrative example, seismic simulation of a real nuclear reactor of High Temperature engineering Test Reactor, which is located at the O-arai research and development center of JAEA, with 80 major mechanical components was conducted. Consequently, our code successfully simulated detailed elasto-plastic deformation of nuclear reactor and its computational performance was investigated. (author)

(E-4-Bromo-N-(2,3-dimethoxybenzylideneaniline

Directory of Open Access Journals (Sweden)

Karla Fejfarová

2010-09-01

Full Text Available The title Schiff base compound, C15H14BrNO2, was prepared by the condensation of 2,3-dimethoxybenzaldehyde with 4-bromoaniline. It adopts an E configuration with respect to the C=N bond. The dihedral angle between the two aromatic rings is 56.79 (8°. Weak C—H...O and C—-H...π bonds can be found in the crystal structure.

Análisis de estabilidad del reactor PFTR para una reacción con cinética de primer orden utilizando la funcional de Lyapunov

OpenAIRE

Héctor Armando Durán Peralta; Luis Fernando Córdoba C

2007-01-01

Abunda la literatura referente al análisis de estabilidad de reactores con parámetros globalizados de concentración y temperatura (por ejemplo el CSTR), en cambio es escasa la literatura sobre la estabilidad de reactores con parámetros distribuidos donde existe distribución espacial de concentración y temperatura, como es el caso del reactor tubular PFTR. Este documento analiza la estabilidad del reactor PFTR isotérmico y no isotérmico para una reacción con cinética de primer orden utilizando...

Análisis de estabilidad del reactor pftr para una reacción con cinética de primer orden utilizando la funcional de lyapunov

OpenAIRE

Durán Peralta, Héctor Armando; Córdoba C, Luis Fernando

2010-01-01

Abunda la literatura referente al análisis de estabilidad de reactores con parámetros globalizados de concentración y temperatura (por ejemplo el CSTR), en cambio es escasa la literatura sobre la estabilidad de reactores con parámetros distribuidos donde existe distribución espacial de concentración y temperatura, como es el caso del reactor tubular PFTR. Este documento analiza la estabilidad del reactor PFTR isotérmico y no isotérmico para una reacción con cinética de primer orden utilizand...

FORMACIÓN DE GRÁNULOS ANAEROBIOS SIN LODOS INOCULADOS EN UN REACTOR DE BIOPELÍCULA ANAEROBIA TIPO INTERCAMBIADOR DE CALOR (RBAIC

Directory of Open Access Journals (Sweden)

Ramiro Escalera Vásquez

2012-07-01

Full Text Available Se ha estudiado experimentalmente el fenómeno de la formación de gránulos anaerobios en la puesta en marcha de un Reactor de Biopelícula Anaerobia Tipo Intercambiador de Calor (RBAIC, utilizado en el tratamiento de aguas residuales provenientes de una planta de procesamiento de melazas de baja concentración, 0,45 g-C L-1. El reactor fue operado en modo de flujo ascendente de un solo paso a bajas temperaturas promedio del líquido (15° C a 25° C y tiempos de retención hidráulica (TRH de 16 h y 6,1 h. La granulación puede proceder sin la inoculación de grandes cantidades iniciales de lodo biológico anaerobio. La granulación es posible aún con concentraciones iniciales muy pequeñas de microorganismos, siempre y cuando los mismos estén bien aclimatados al tipo de agua residual y las condiciones ambientales. Es posible tratar tasas de carga orgánica volumétrica entre 1,8 y 3,3 g-DQO L-1 d-1 a temperaturas promedio menores a 25° C con concentraciones de biomasa tan pequeñas como 2 g-SSV L-1. En un RBAIC, se pueden formar gránulos de 0,5 – 2 mm de diámetro en tres meses a temperaturas promedio menores a 25° C.

Investigation of (n,γ) reaction in hybrid reactor zones

International Nuclear Information System (INIS)

Guenay, Mehtap

2014-01-01

In this study, the fluids were composed with increased mole fractions of a mixture of molten salt: heavy metals 99-95 % Li 20 Sn 80 - 1-5 % SFG-Pu, 99-95 % Li 20 Sn 80 - 1-5 % SFG-PuF 4 , 99-95 % Li 20 Sn 80 - 1-5 % SFG-PuO 2 . In this study, the effect on conversion of each isotope ( 238-242 Pu) in spent fuel grade plutonium by (n,γ) reactions was investigated in liquid first wall, blanket and shield zones of the designed hybrid reactor system. Beryllium (Be) is the neutron multiplier by (n,2n) reactions. The Be zone used was 3 cm thick. 9Cr2WVT, a ferritic steel, is used as structural material. Three-dimensional nucleonic calculations were performed by using the most recent versions of the MCNPX-2.7.0 Monte Carlo code and the nuclear data library ENDF/B-VII.0.

Investigation of (n,γ) reaction in hybrid reactor zones

Energy Technology Data Exchange (ETDEWEB)

Guenay, Mehtap [Inoenue Univ., Malatya (Turkey). Physics Dept.

2014-12-15

In this study, the fluids were composed with increased mole fractions of a mixture of molten salt: heavy metals 99-95 % Li{sub 20}Sn{sub 80{sup -}}1-5 % SFG-Pu, 99-95 % Li{sub 20}Sn{sub 80{sup -}}1-5 % SFG-PuF{sub 4}, 99-95 % Li{sub 20}Sn{sub 80{sup -}}1-5 % SFG-PuO{sub 2}. In this study, the effect on conversion of each isotope ({sup 238-242}Pu) in spent fuel grade plutonium by (n,γ) reactions was investigated in liquid first wall, blanket and shield zones of the designed hybrid reactor system. Beryllium (Be) is the neutron multiplier by (n,2n) reactions. The Be zone used was 3 cm thick. 9Cr2WVT, a ferritic steel, is used as structural material. Three-dimensional nucleonic calculations were performed by using the most recent versions of the MCNPX-2.7.0 Monte Carlo code and the nuclear data library ENDF/B-VII.0.

Características da sucção não-nutritiva em RN a termo e pré-termo tardio Characteristics of non-nutritive sucking in full-term and late preterm infants

Directory of Open Access Journals (Sweden)

Ana Paula d'Oliveira Gheti Kao

2011-09-01

Full Text Available OBJETIVO: Comparar os parâmetros de sucção não nutritiva de recém-nascidos a termo e pré-termo tardio. MÉTODOS: Os recém-nascidos foram divididos em dois grupos, pré-termo tardio (RNPT tardio e a termo (RN a termo e, submetidos à avaliação da sucção não-nutritiva utilizando-se um protocolo adaptado da Escala de Avaliação Motora Oral. Foi realizada análise estatística para comparação dos grupos. RESULTADOS: Os reflexos de procura e de sucção foram menos frequentes nos RNPT tardio, comparados aos RN a termo, assim como a preensão palmar e mãos em linha média. A maioria dos RNPT tardio apresentou sono leve ou estava sonolento antes da avaliação. Os RNPT tardio apresentaram predominantemente sucção esporádica ou grupos de sucção com pausas longas e travamento e/ou tremores de mandíbula. A retração de língua e a protrusão de língua foram mais presentes nos RNPT tardio e o canolamento de língua nos RN a termo. CONCLUSÃO: Prontidão para a mamada, estado comportamental, postura corporal, padrão e força de sucção e movimentos de língua foram os parâmetros menos frequentes nos RNPT tardio em relação aos RN a termo.PURPOSE: To compare non-nutritive sucking parameters between late preterm and full-term infants. METHODS: Infants were divided into two groups, full-term and late preterm, and were submitted to non-nutritive sucking assessment using a protocol adapted from the Oral Motor Assessment Scale. Statistical analysis was conducted for comparison between the groups. RESULTS: The seeking and sucking reflexes were less frequent in late preterm than in full-term newborns, as well as palmar grip and hands in the midline. Most late preterm infants presented light sleep or drowsiness before the assessment. Late preterm subjects predominantly presented sporadic sucking or blocks of sucking with long pauses and mandibular locking and/or tremors. Tongue retraction and protrusion were mostly present in late preterm

Hanford Site 100-N Area In Situ Bioremediation of UPR-100-N-17, Deep Petroleum Unplanned Release - 13245

Energy Technology Data Exchange (ETDEWEB)

Saueressig, Daniel G. [Washington Closure Hanford, 2620 Fermi, Richland, Washington, 99354 (United States)

2013-07-01

In 1965 and 1966, approximately 303 m{sup 3} of Number 2 diesel fuel leaked from a pipeline used to support reactor operations at the Hanford Site's N Reactor. N Reactor was Hanford's longest operating reactor and served as the world's first dual purpose reactor for military and power production needs. The Interim Action Record of Decision for the 100-N Area identified in situ bioremediation as the preferred alternative to remediate the deep vadose zone contaminated by this release. A pilot project supplied oxygen into the vadose zone to stimulate microbial activity in the soil. The project monitored respiration rates as an indicator of active biodegradation. Based on pilot study results, a full-scale system is being constructed and installed to remediate the vadose zone contamination. (authors)

Emergency response guide-B ECCS guideline evaluation analyses for N reactor

International Nuclear Information System (INIS)

Chapman, J.C.; Callow, R.A.

1989-07-01

INEL conducted two ECCS analyses for Westinghouse Hanford. Both analyses will assist in the evaluation of proposed changes to the N Reactor Emergency Response Guide-B (ERG-B) Emergency Core System (ECCS) guideline. The analyses were a sensitivity study for reduced-ECCS flow rates and a mechanistically determined confinement steam source for a delayed-ECCS LOCA sequence. The reduced-ECCS sensitivity study established the maximum allowable reduction in ECCS flow as a function of time after core refill for a large break loss-of-coolant accident (LOCA) sequence in the N Reactor. The maximum allowable ECCS flow reduction is defined as the maximum flow reduction for which ECCS continues to provide adequate core cooling. The delayed-ECCS analysis established the liquid and steam break flows and enthalpies during the reflood of a hot core following a delayed ECCS injection LOCA sequence. A simulation of a large, hot leg manifold break with a seven-minute ECCS injection delay was used as a representative LOCA sequence. Both analyses were perform using the RELAP5/MOD2.5 transient computer code. 13 refs., 17 figs., 3 tabs

Experimental cross section for the {sup 152}Sm(n, γ){sup 153}Sm reaction at 0.0334 eV

Energy Technology Data Exchange (ETDEWEB)

Uddin, M. Shuza; Datta, Tapash Kumar; Hossain, Syed Mohammod; Zakaria, A.K.M.; Islam, Mohammad Amirul; Naher, Kamrun; Shariff, M. Asad; Yunus, S.M. [Atomic Energy Research Establishment, Dhaka (Bangladesh). Inst. of Nuclear Science and Technology; Afroze, Nasmin [Atomic Energy Research Establishment, Dhaka (Bangladesh). Inst. of Nuclear Science and Technology; Jahangirnagar Univ., Dhaka (Bangladesh). Dept. of Physics; Islam, S.M. Ajharul [Jahangirnagar Univ., Dhaka (Bangladesh). Dept. of Physics

2014-10-01

The neutron capture cross section for the {sup 152}Sm(n, γ){sup 153}Sm reaction at an energy of 0.0334 eV was measured for the first time using monochromatic neutrons of a powder diffractometer at the TRIGA Mark II nuclear reactor at Dhaka, Bangladesh. The {sup 197}Au(n, γ){sup 198}Au reaction was used to monitor the neutron beam intensity. The radioactivity of the products was determined via high resolution γ-ray spectrometry. The obtained cross section value is 184 ± 22b, which is consistent with both the ENDF/B-VII and TENDL-2012 data libraries. The measured value at 0.0334 eV and the previous data at 0.0536 eV confirm the reliability of the data in the above libraries. (orig.)

Pronunciación y TIC en el aula de E/LE

Directory of Open Access Journals (Sweden)

Nuria Escrig Teixeira

2016-11-01

Full Text Available El objetivo de la siguiente experiencia práctica es presentar y describir dos aplicaciones con las que trabajar la pronunciación en el aula de E/LE: Adobe Voice y Dragon Dictation. Mediante estas dos herramientas podemos desarrollar las cuatro habilidades de la lengua (comprensión lectora, expresión escrita, comprensión oral y expresión oral, aplicar los contenidos explicados en cualquier nivel, introducir las TIC en clase y practicar la pronunciación, materia un poco olvidada en el aula. Por ello, se realizará una descripción de las aplicaciones y de las posibilidades didácticas para explotar su uso en clase.   Palabras clave: aplicaciones, pronunciación, E/LE, posibilidades didácticas.   Pronunciation and ICT in E/LE class   Abstract: The objetive of the following experience is to introduce and describe two apps to work the pronunciation in E/LE class: Adobe Voice and Dragon Dictation. Through this apps the four comuniucation skills (speaking, listening, writing and reading are developed in any level and the ICT are introduced in a way we can practise the pronunciation, forgotten subject in class. For that reason, we will describe the apps and their didactic possibilities to use.   Keywords: apps, pronunciation, E/LE, didactic possibilities.

El docente ante la evaluación de la expresión e interacción orales en ELE

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López Moya, Consuelo

2015-05-01

Full Text Available Aunque cada vez más la expresión y la interacción orales están siendo dotadas de mayor relevancia en la clase de español como lengua extranjera, su evaluación todavía escapa de una sistematización conjunta en dicha enseñanza. Ante esta situación, el presente artículo trata de acercarse a esa realidad y, para ello, profesores de ELE de diferentes centros han mostrado tanto su modo de proceder en la evaluación de estas destrezas como el propio de su centro. A través de un cuestionario ha sido valorado este proceso desde una triple vertiente: considerando el grado de importancia otorgado a la expresión e interacción orales, según el método desarrollado para evaluar estas destrezas y atendiendo al proceso de evaluación desde la óptica del docente. Las respuestas han reafirmado la inexistencia de un consenso a la hora de llevar a cabo la evaluación de la expresión e interacción orales en la enseñanza del español como lengua extranjera.

La e-investigación de la Comunicación: actitudes, herramientas y prácticas en investigadores iberoamericanos

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Carlos Arcila Calderón

2013-03-01

Full Text Available La e-investigación está cambiando las prácticas y dinámicas de la investigación social, gracias a la incorporación de herramientas digitales avanzadas para el procesamiento de datos y el incremento de la colaboración científica. Estudios anteriores muestran una actitud positiva de los científicos hacia la e-investigación y la rápida incorporación de herramientas digitales para el trabajo académico, a pesar de las resistencias culturales al cambio. Este artículo examina el estado actual (actitudes, herramientas y prácticas de la e-investigación en el campo de los estudios en comunicación en Iberoamérica. Un total de 316 investigadores de la región respondieron una encuesta en línea durante los últimos dos meses de 2011. Los resultados confirman una actitud positiva hacia la e-investigación y un uso frecuente de las e-herramientas. Sin embargo, la mayor parte de ellos aseguran usar e-herramientas básicas (como correo-e, videoconferencia comercial, software de oficina o redes sociales, en vez de usar tecnologías avanzadas para procesar gran cantidad de datos (como Grids, programas de simulación o Internet2 o de incorporarse a comunidades virtuales de investigación. Algunos investigadores afirmaron tener un uso «intensivo» (31% o «frecuente» (53% de las e-herramientas, pero solo el 22% aseguraron que la capacidad de su computador personal era insuficiente para manejar y procesar los datos. El artículo concluye evidenciando una brecha importante entre la e-investigación en comunicación y en otras disciplinas, y establece recomendaciones para su implementación en la región.

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

Power balance in an Ohmically heated fusion reactor

International Nuclear Information System (INIS)

Christiansen, J.P.; Roberts, K.V.

1982-01-01

A simplified power-balance equation (zero-dimensional model) is used to study the performance of an Ohmically heated fusion reactor with emphasis on a pulsed reversed-field pinch concept (RFP). The energy confinement time tausub(E) is treated as an adjustable function, and empirical tokamak scaling laws are employed in the numerical estimates, which are supplemented by 1-D ATHENE code calculations. The known heating rates and energy losses are represented by the net energy replacement time tausub(W), which is exhibited as a surface in density (n) and temperature (T) space with a saddle point (nsub(*), Tsub(*)), the optimum ignition point. It is concluded that i) ignition by Ohmic heating is more practicable for the RFP reactor than for a tokamak reactor with the same tausub(E), (ii) if at fixed current the minor radius can be reduced or at fixed minor radius the current can be increased, then it is found that Ohmic ignition becomes more likely when present tokamak scaling laws are used. More definitive estimates require, however, a knowledge of tausub(E), which can only be obtained by establishing a reliable set of experimental RFP scaling laws and, in particular, by extending RFP experiments closer to the reactor regime. (author)

El registro civil e identificación en la región y el mundo

Directory of Open Access Journals (Sweden)

Felix Ortega de la Torre

2018-01-01

Full Text Available Realiza un repaso histórico, así como un recorrido por el desarrollo del proceso registral civil en la región latinoamericana. Enfatiza en la importancia que existe en la relación entre el registro de personas y el desarrollo, la democracia, la gobernabilidad; así como para la inclusión, la transparencia, la seguridad, la eficiencia de recursos y la generación de un Estado moderno digital. Puntualmente, analiza los modelos de registro e identificación, el fortalecimiento del registro civil en el siglo veintiuno, las tendencias futuras, el registro universal y sus nuevos desafíos

Evaluación del comportamiento hidráulico en un reactor anaerobio de doble cámara (RADCA)

OpenAIRE

Nancy Rincón; Andres Galindo; Jhonny Pérez

2011-01-01

En la remoción de carga orgánica en un sistema de tratamiento de aguas residuales intervienen los procesos bioquímico y aspectos hidrodinámicos como las características de flujo, régimen de mezcla, tiempos de residencia, geometría del reactor, por otro lado las condiciones de flujo no ideal tales como cortos circuitos, zonas muertas y recirculación interna afectan su desempeño. En esta investigación se evaluó el comportamiento hidráulico de un reactor anaerobio de doble cámara (RADCA) de 5...

E Günıster

Indian Academy of Sciences (India)

Home; Journals; Bulletin of Materials Science. E Günıster. Articles written in Bulletin of Materials Science. Volume 27 Issue 3 June 2004 pp 317-322 Surfactants. Effect of sodium dodecyl sulfate on flow and electrokinetic properties of Na-activated bentonite dispersions · E Günıster S İşçı A Alemdar N Güngör · More Details ...

Méthodologie de l'extrapolation des réacteurs chimiques Methodology for Scaling Up Chemical Reactors

Directory of Open Access Journals (Sweden)

Trambouze P.

2006-11-01

Full Text Available Après un exposé général relatif à la méthodologie du développement des procédés, applicable à l'extrapolation des réacteurs, est présenté un rapide examen critique des deux principales techniques mises en oeuvre, à savoir : - la théorie de la similitude ; - l'élaboration de modèles mathématiques. Deux exemples pratiques, relatifs aux réacteurs homogènes et aux réacteurs catalytiques à lit fixe et deux phases fluides, sont ensuite examinés à la lumière des considérations générales précédentes. After giving a general description of process-development methodology applicable to scaling up reactors, this article makes a quick critical examination of the two main techniques involved, i. e. : (a the theory of similarity, and (b the compiling of mathematical models. Two practical examples relating to homogeneous reactors and trickle-bed catalytic reactors are then examined in the light of the preceding general considerations.

Sistema multi-robot para localización e identificación de vehículos

Directory of Open Access Journals (Sweden)

C. Sagues

2012-01-01

Full Text Available Resumen: En este trabajo se presenta un sistema multi-robot para localización e identificación de vehículos que están estacionados en un recinto abierto o cerrado. El sistema realiza una planificación a priori y una asignación de tareas a los miembros del equipo optimizando el tiempo de la misión. El equipo de robots está dotado de sensores de visión que permiten la localización de los vehículos y la identificación de su matrícula. El controlador de movimiento de cada robot utiliza un sensor láser para el posicionamiento frente al vehículo a identificar y un sistema de control basado en visión realiza el posicionamiento preciso para la adquisición de la imagen de la matrícula que permita su posterior identificación. El sistema multi-robot dispone de capacidad de comunicaciones entre ellos y con una estación central de mando, con la que se intercambian comandos e incidencias y eventualmente datos, con restricciones de tiempo real. Los sensores utilizados están comercialmente disponibles y los algoritmos han sido desarrollados por el grupo Robótica, Percepción y Tiempo Real de la Universidad de Zaragoza en el marco de proyectos financiados por el Ministerio de Ciencia e Innovación. Integra diversas tecnologías de planificación, navegación, percepción y comunicaciones, adaptadas en el proyecto a la aplicación concreta. Palabras clave: Sistemas multi-robot, Robótica móvil, Navegación, Planificación y asignación de tareas, Protocolos de comunicación, Visión porcomputador, Tiempo real

Ningyo Toge uranium enrichment pilot plant comes into full

International Nuclear Information System (INIS)

1982-01-01

The uranium enrichment pilot plant of the Power Reactor and Nuclear Fuel Development Corporation at Ningyo Toge went into full operation on March 26, 1982. This signifies that the front end of the nuclear fuel cycle in Japan, from uranium ore to enrichment, is only a step away from commercialization. On the same day, the pilot plant of uranium processing and conversion to UF 6 , the direct purification of uranium ore into uranium hexafluoride, began batch operation at the same works. The construction of the uranium enrichment pilot plant has been advanced in three stages: i.e. OP-1A with 1000 centrifuges, OP-1B with 3000 centrifuges and OP-2 with 3000 centrifuges. With a total of 7000 centrifuges, the pilot plant, the first enrichment plant in Japan, has now a capacity of supplying enriched uranium for six months operation of a 1,000 MW nuclear power plant. (J.P.N.)

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sato, Morihiko.

1994-01-01

Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

Very high flux steady state reactor and accelerator based sources

International Nuclear Information System (INIS)

Ludewig, H.; Todosow, M.; Simos, N.; Shapiro, S.; Hastings, J.

2004-01-01

With the number of steady state neutron sources in the US declining (including the demise of the Bnl HFBR) the remaining intense sources are now in Europe (i.e. reactors - ILL and FMR, accelerator - PSI). The intensity of the undisturbed thermal flux for sources currently in operation ranges from 10 14 n/cm 2 *s to 10 15 n/cm 2 *s. The proposed Advanced Neutron Source (ANS) was to be a high power reactor (about 350 MW) with a projected undisturbed thermal flux of 7*10 15 n/cm 2 *s but never materialized. The objective of the current study is to explore the requirements and implications of two source concepts with an undisturbed flux of 10 16 n/cm 2 *s. The first is a reactor based concept operating at high power density (10 MW/l - 15 MW/l) and a total power of 100 MW - 250 MW, depending on fissile enrichment. The second is an accelerator based concept relying on a 1 GeV - 1.5 GeV proton Linac with a total beam power of 40 MW and a liquid lead-bismuth eutectic target. In the reactor source study, the effects of fissile material enrichment, coolant temperature and pressure drop, and estimates of pressure vessel stress levels will be investigated. The fuel form for the reactor will be different from all other operating source reactors in that it is proposed to use an infiltrated graphitic structure, which has been developed for nuclear thermal propulsion reactor applications. In the accelerator based source the generation of spallation products and their activation levels, and the material damage sustained by the beam window will be investigated. (authors)

Neutronics conceptual design of the innovative research reactor core using uranium molybdenum fuel

International Nuclear Information System (INIS)

Tukiran S; Surian Pinem; Tagor MS; Lily S; Jati Susilo

2012-01-01

The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research newest. Reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR) and then as an alternative to replace the old research reactor. The aim of this research is to get the optimal configuration of equilibrium core with the acceptance criteria are minimum thermal neutron flux is 2.5E14 n/cm 2 s at the power level of 20 MW (minimum), length of cycle of more than 40 days, and the most efficient of using fuel in the core. Neutronics design has been performed for new fuel of U-9Mo-AI with various fuel density and reflector. Design calculation has been performed using WIMSD-5B and BATAN-FUEL computer codes. The calculation result of the conceptual design shows four core configurations namely 5x5, 5x7, 6x5 and 6x6. The optimalization result for equilibrium core of innovative research reactor is the 5x5 configuration with 450 gU fuel loading, berilium reflector, maximum thermal neutron flux at reflector is 3.33E14 n/cm 2 sand length of cycle is 57 days is the most optimal of IRR. (author)

COMUNICACIÓN GLOBAL Y CAPACIDAD COMPETITIVA EN LAS MICRO EMPRESAS DE COMUNICACIÓN E INFORMACIÓN

Directory of Open Access Journals (Sweden)

Ángel Páez

2001-01-01

Full Text Available La presente investigación ahonda en la caracterización de las micro empresas de comunicación e información (MECI, empresas que han aparecido en el ámbito mundial a raíz del crecimiento y expansión de las tecnologías de información y comunicación. Se explica en esta investigación la relación entre la comunicación global y la competitividad de las MECI de Maracaibo. Es una investigación de tipo correlacional. Para la recolección de datos se utilizó la técnica de la entrevista estruturada, efectuándola al personal gerencial, administrativo y operativo de las MECI. Los resultados obtenidos comprueban la hipótesis de que existe una relación entre la comunicación global y la capacidad competitiva. El trabajo constituye una contribución para las MECI, debido a que su estudio sistemático proporciona una serie de criterios teóricos y prácticos que podrían ser utilizados por sus gerentes para generar competitividad en el futuro.

Spectrometer requirements for (e,e'2N) studies

International Nuclear Information System (INIS)

Lightbody, J.W. Jr.

1981-01-01

One specific experiment that may be performed with a future CW accelerator is a study of (e,e'2N) reactions through which we may learn details of the short range interaction of two nucleons within nuclear matter. It is suspected that the only mechanism which can lead to the observed high momentum components in the single nucleon momentum distribution (above approx. 400 MeV/c) inferred from (e,e'p) and (γ,p) measurements is the presence of short-range few-body correlations in the many-body nuclear wave function. It is expected that the explicit pair correlation function may be inferred from relative two-nucleon momentum distributions measured in (e,e'2N) experiments. It is therefore interesting to estimate counting rates using measured one-body momentum distributions to see what types of spectrometers are required

SOLUCIÓN ANALÍTICA PARA OBTENER EL VOLUMEN ÓPTIMO DE UNA SERIE DE REACTORES DE AGITACIÓN CONTINUA DONDE SE EFECTÚA UNA REACCIÓN DE PRIMER ORDEN

Directory of Open Access Journals (Sweden)

Ignacio Elizalde

2013-01-01

Full Text Available An analytical procedure for determining the optimum size of CSTR in series operating under isothermal and isobaric conditions sustaining first order reaction at constant density has been developed. The procedure requires the concentration of reactant at the entrance of the first reactor and at the outlet of the last reactor; it is also required the continuity of reaction rate as function of conversion, due to the later changes from one reactor to another. The optimization method involves the calculation of intermediate concentrations instead of their estimation, as it is done by graphical solution reported previously. Also, the procedure reported in this contribution is valid for any reactor number. Under these circumstances the method predicts that all reactors must have the same size in order to minimize the total volume of the system.

LA ADOPCIÓN DEL E-GOBIERNO EN ENTORNOS VOLUNTARIOS

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Medina Molina, Cayetano

2013-01-01

Full Text Available El desarrollo del e-Gobierno centrado en el usuario requiere conocer tanto las expectativas que este tiene respecto a su uso, como los elementos que favorecen su adopción. Chan et al. (2010 han planteado la existencia de diversos antecedentes de los componentes del modelo Unified Theory of Acceptance and Use of Technology (UTAUT en un entorno obligatorio. El presente trabajo, a través del desarrollo de un modelo de ecuaciones estructurales mediante PLS, analiza la vigencia de tales antecedentes en un entorno voluntario. Los resultados muestran cómo las expectativas de resultado y las expectativas de esfuerzo inciden de forma significativa sobre la intención de emplear la plataforma de e-Gobierno, y también que esta y las condiciones facilitadoras influyen sobre el uso de la mencionada plataforma. Respecto a los antecedentes de los componentes del modelo UTAUT, cabe destacar el papel desempe˜nado por la conveniencia, asistencia, confianza y aversión a la interacción personal.

Characteristics of a novel nanosecond DBD microplasma reactor for flow applications

Science.gov (United States)

Elkholy, A.; Nijdam, S.; van Veldhuizen, E.; Dam, N.; van Oijen, J.; Ebert, U.; de Goey, L. Philip H.

2018-05-01

We present a novel microplasma flow reactor using a dielectric barrier discharge (DBD) driven by repetitive nanosecond high-voltage pulses. Our DBD-based geometry can generate a non-thermal plasma discharge at atmospheric pressure and below in a regular pattern of micro-channels. This reactor can work continuously up to about 100 min in air, depending on the pulse repetition rate and operating pressure. We here present the geometry and main characteristics of the reactor. Pulse energies of 1.46 and 1.3 μJ per channel at atmospheric pressure and 50 mbar, respectively, have been determined by time-resolved measurements of current and voltage. Time-resolved optical emission spectroscopy measurements have been performed to calculate the relative species concentrations and temperatures (vibrational and rotational) of the discharge. The effects of the operating pressure and flow velocity on the discharge intensity have been investigated. In addition, the effective reduced electric field strength {(E/N)}eff} has been obtained from the intensity ratio of vibronic emission bands of molecular nitrogen at different operating pressures and different locations. The derived {(E/N)}eff} increases gradually from about 550 to 4600 Td when decreasing the pressure from 1 bar to 100 mbar. Below 100 mbar, further pressure reduction results in a significant increase in {(E/N)}eff} up to about 10000 Td at 50 mbar.

FACTORES DE CRECIMIENTO EN EL DESARROLLO FOLICULAR, EMBRIONARIO TEMPRANO E IMPLANTACIÓN. IMPLICACIONES EN LA PRODUCCIÓN DE EMBRIONES BOVINOS

Directory of Open Access Journals (Sweden)

Miguel Peña J

2007-08-01

Full Text Available Se describe el papel de los factores de crecimiento en el desarrollo folicular, embrionariotemprano e implantación y se destaca su importancia en la producción de embrionesbovinos. Los factores de crecimiento cumplen importantes funciones en eventos clavesde la actividad reproductiva. Aunque se conocen algunos mecanismos de acción y elefecto sinérgico con otras hormonas reproductivas, se está lejos de conocer totalmentesu papel. Estudios previos tanto in vivo como in vitro señalan que podrían mejorarnotablemente las tasas de fertilización, de clivaje e implantación.

White emission by self-regulated growth of InGaN/GaN quantum wells on in situ self-organized faceted n-GaN islands

International Nuclear Information System (INIS)

Fang Zhilai

2011-01-01

The in situ self-organization of three-dimensional n-GaN islands of distinct sidewall faceting was realized by initial low V/III ratio growth under high reactor pressure followed by variations of the V/III ratio and reactor pressure. The naturally formed faceted islands with top and sidewall facets of various specific polar angles may serve as an ideal template for self-regulated growth of the InGaN/GaN multiple quantum wells (MQWs), i.e. the growth behavior is specific polar angle dependent. Further, the growth behavior and luminescence properties of the InGaN/GaN MQWs on various facets of different specific polar angles are directly compared and discussed. Tetrachromatic white emissions (blue, cyan, green, and red) from single-chip phosphor-free InGaN/GaN MQWs are realized by color tuning through island shaping, shape variations, and self-regulated growth of the InGaN/GaN MQWs.

Performance evaluation of 24 ion exchange materials for removing cesium and strontium from actual and simulated N-Reactor storage basin water

Energy Technology Data Exchange (ETDEWEB)

Brown, G.N.; Carson, K.J.; DesChane, J.R.; Elovich, R.J.

1997-09-01

This report describes the evaluation of 24 organic and inorganic ion exchange materials for removing cesium and strontium from actual and simulated waters from the 100 Area 105 N-Reactor fuel storage basin. The data described in this report can be applied for developing and evaluating ion exchange pre-treatment process flowsheets. Cesium and strontium batch distribution ratios (K{sub d}`s), decontamination factors (DF), and material loadings (mmol g{sup -1}) are compared as a function of ion exchange material and initial cesium concentration. The actual and simulated N-Basin waters contain relatively low levels of aluminum, barium, calcium, potassium, and magnesium (ranging from 8.33E-04 to 6.40E-05 M), with slightly higher levels of boron (6.63E-03 M) and sodium (1.62E-03 M). The {sup 137}Cs level is 1.74E-06 Ci L-{sup 1} which corresponds to approximately 4.87E-10 M Cs. The initial Na/Cs ratio was 3.33E+06. The concentration of total strontium is 4.45E-06 M, while the {sup 90}Sr radioactive component was measured to be 6.13E-06 Ci L{sup -1}. Simulant tests were conducted by contacting 0.067 g or each ion exchange material with approximately 100 mL of either the actual or simulated N-Basin water. The simulants contained variable initial cesium concentrations ranging from 1.00E-04 to 2.57E- 10 M Cs while all other components were held constant. For all materials, the average cesium K{sub d} was independent of cesium concentration below approximately 1.0E-06 M. Above this level, the average cesium K{sub d} values decreased significantly. Cesium K{sub d} values exceeding 1.0E+07 mL g{sup -1} were measured in the simulated N-Basin water. However, when measured in the actual N-Basin water the values were several orders of magnitude lower, with a maximum of 1.24E+05 mL g{sup -1} observed.

Nuclear reactor

International Nuclear Information System (INIS)

Schulze, I.; Gutscher, E.

1980-01-01

The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de

TECNOLOGÍAS PARA LA DECOLORACIÓN DE TINTES ÍNDIGO E ÍNDIGO CARMÍN

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LUZ QUINTERO

2010-01-01

Full Text Available Los tratamientos de agua residual textil con los tintes índigo e índigo carmín son muy complejos y variados. La eficiencia varía de acuerdo al método empleado. Este artículo revisa las diferentes tecnologías de tratamiento para la remoción de índigo e índigo carmín; eficiencias de remoción, cultivos empleados, sistemas de procesos, factores operacionales, entre otros aspectos, con el fin de establecer criterios para seleccionar el mejor proceso de tratamiento y conocer los alcances de la investigación en la decoloración de índigo. La revisión comienza con el proceso de fijación en el teñido de índigo y luego describe los estudios de tratamiento de efluente con índigo a escala laboratorio y a gran escala. Existen tecnologías físicoquímicas, químicas, físicas y biológicas para la decoloración de agua con índigo. La elección del tratamiento está en función de la calidad de agua del efluente, el uso, los costos de la tecnología, ventajas y desventajas.

Low drift type N thermocouples in out-of-pile advanced gas reactor mock-up test: metallurgical analysis

International Nuclear Information System (INIS)

Scervini, M.; Palmer, J.; Haggard, D.C.; Swank, W.D.

2015-01-01

Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. They are crucial for the control of current nuclear reactors and for the development of GEN IV reactors. In nuclear applications thermocouples are strongly affected by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. Previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of Nickel based thermocouples is limited to temperatures lower than 1000 deg. C due to drift related to phenomena other than nuclear irradiation. As part of a collaboration between Idaho National Laboratory (INL) and the University of Cambridge a variety of Type N thermocouples have been exposed at INL in an Advanced Gas Reactor mock-up test at 1150 deg. C for 2000 h, 1200 deg. C for 2000 h, 125 deg. C for 200 h and 1300 deg. C for 200 h, and later analysed metallurgically at the University of Cambridge. The use of electron microscopy allows to identify the metallurgical changes occurring in the thermocouples during high temperature exposure and correlate the time dependent thermocouple drift with the microscopic changes experienced by the thermoelements of different thermocouple designs. In this paper conventional Inconel 600 sheathed type N thermocouples and a type N using a customized sheath developed at the University of

Low drift type N thermocouples in out-of-pile advanced gas reactor mock-up test: metallurgical analysis

Energy Technology Data Exchange (ETDEWEB)

Scervini, M. [University of Cambridge, Department of Materials Science and Metallurgy, 27 Charles Babbage Road, CB30FS Cambridge, (United Kingdom); Palmer, J.; Haggard, D.C.; Swank, W.D. [Idaho National Laboratory, Idaho Falls, ID 83415-3840, (United States)

2015-07-01

Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. They are crucial for the control of current nuclear reactors and for the development of GEN IV reactors. In nuclear applications thermocouples are strongly affected by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. Previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of Nickel based thermocouples is limited to temperatures lower than 1000 deg. C due to drift related to phenomena other than nuclear irradiation. As part of a collaboration between Idaho National Laboratory (INL) and the University of Cambridge a variety of Type N thermocouples have been exposed at INL in an Advanced Gas Reactor mock-up test at 1150 deg. C for 2000 h, 1200 deg. C for 2000 h, 125 deg. C for 200 h and 1300 deg. C for 200 h, and later analysed metallurgically at the University of Cambridge. The use of electron microscopy allows to identify the metallurgical changes occurring in the thermocouples during high temperature exposure and correlate the time dependent thermocouple drift with the microscopic changes experienced by the thermoelements of different thermocouple designs. In this paper conventional Inconel 600 sheathed type N thermocouples and a type N using a customized sheath developed at the University of

Flowsheet for shear/leach processing of N Reactor fuel at PUREX

International Nuclear Information System (INIS)

Enghusen, M.B.

1995-01-01

This document was originally prepared to support the restart of the PUREX plant using a new Shear/Leach head end process. However, the PUREX facility was shutdown and processing of the remaining N Reactor fuel is no longer considered an alternative for fuel disposition. This document is being issued for reference only to document the activities which were investigated to incorporate the shear/leach process in the PUREX plant

Neutron beam facilities at the Australian Replacement Research Reactor

International Nuclear Information System (INIS)

Kennedy, Shane; Robinson, Robert; Hunter, Brett

2001-01-01

Australia is building a research reactor to replace the HIFAR reactor at Lucas Heights by the end of 2005. Like HIFAR, the Replacement Research Reactor will be multipurpose with capabilities for both neutron beam research and radioisotope production. It will be a pool-type reactor with thermal neutron flux (unperturbed) of 4 x 10 14 n/cm 2 /sec and a liquid D 2 cold neutron source. Cold and thermal neutron beams for neutron beam research will be provided at the reactor face and in a large neutron guide hall. Supermirror neutron guides will transport cold and thermal neutrons to the guide hall. The reactor and the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP S.E. under contract. The neutron beam instruments will be developed by ANSTO, in consultation with the Australian user community. This status report includes a review the planned scientific capabilities, a description of the facility and a summary of progress to date. (author)

Embrittlement of the Shippingport reactor shield tank

International Nuclear Information System (INIS)

Chopra, O.K.; Shack, W.J.

1989-01-01

Surveillance specimens from the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory showed an unexpectedly high degree of embrittlement relative to the data obtained on similar materials in Materials Testing Reactors (MTRs). The results suggest a possible negative flux effect and raise the issue of embrittlement of the pressure vessel support structures of commercial light water reactors. To help resolve this issues, a program was initiated to characterize the irradiation embrittlement of the neutron shield tank (NST) from the decommissioned Shippingport reactor. The Shippingport NST operated at 55 degree C (130 degree F) and was fabricated from A212 Grade B steel, similar to the vessel material in HFIR. The inner wall of the NST was exposed to a total maximum fluence of ∼ 6 x 10 17 n/cm 2 (E > 1 MeV) over a life of 9.25 effective full power years. This corresponds to a fast flux of 2.1 x 10 9 n/cm 2 x s and is comparable to the conditions for the HFIR surveillance specimens. The results indicate that irradiation increases the 15 ft x lb Charpy transition temperature (CTT) by ∼25 degree C (45 degree F) and decreases the upper shelf energy. The shift in CTT is not as severe as that observed in the HFIR surveillance specimens and is consistent with that expected from the MTR data base. However, the actual value of CTT is high, and the toughness at service temperature is low, even when compared with the HFIR data. The increase in yield stress is ∼50 MPa, which is comparable to the HFIR data. The results also indicate a lower impact strength and higher transition temperature for the TL orientation than that for the LT orientation. Some effects of the location across the thickness of the wall are also observed for the LT specimens; CTT is slightly greater for the specimens from the inner region of the wall

Comunicación e intercambio con redes sociales en la educación universitaria: caso estudiantes de Administración e Informática

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Rodrigo Sandoval-Almazán

2013-10-01

Full Text Available El uso de las redes sociales entre estudiantes se ha extendido a partir del abaratamiento de los costos, el uso de la telefonía celular y el efecto viral de las plataformas de Twitter y Facebook a escala mundial. Sin embargo, apenas se comienza a estudiar su impacto tanto en los procesos de aprendizaje como en los mecanismos de comunicación entre los alumnos. Esta investigación se centra en esta segunda opción, intentando responder la interrogante: ¿cómo usan los estudiantes universitarios las redes sociales para mejorar su comunicación e intercambiar conocimientos? A partir de la aplicación de dos encuestas en línea aplicadas en 2010 y 2012, tanto a estudiantes de Informática Administrativa como de Administración, se logró observar que los alumnos de Administración son los que más utilizan estas plataformas para intercambiar información y que, en un par de años, el uso de estas herramientas ha cambiado drásticamente entre los estudiantes universitarios.

Neutronic analysis concerning the utilization of mixed U N-Pu N nitride fuel for fast reactors

International Nuclear Information System (INIS)

Renke, C.A.C.; Batista, J.L.; Waintraub, M.; Santos Bastos, W. dos; Brito Aghina, L.O. de.

1991-08-01

Neutronic behavior of mixed UN-PuN nitride fuel in substitution of the mixed oxide U O 2 - Pu O 2 for fast reactors is discussed with focus on Super Phenix I. Characteristics parameters of both cores are calculated and compared and the results presented show a great advantage for the nitride fuel, pointing out a larger performance of fuel elements in the core and an effective reduction of reactivity loss during the cycle. (author)

Utility requirements for advanced light water reactors

International Nuclear Information System (INIS)

Machiels, A.; Gray, S.; Mulford, T.; Rodwell, E.

1996-01-01

The nuclear energy industry is actively engaged in developing advanced light water reactor (ALWR) designs for the next century. The new designs take advantage of the thousands of reactor-years of experience that have been accumulated by operating over 400 plants worldwide. The EPRI effort began in the early 1980's, when a survey of utility executives was conducted to determine their prerequisites for ordering nuclear power plants. The results were clear: new plants had to be simpler and safer, and have greater design margins, i.e., be more forgiving. The utility executives also supported making improvements to the established light water reactor technology, rather than trying to develop new reactor concepts. Finally, they wanted the option to build mid-size plants (∼600 MWe) in addition to full-size plants of more than 1200 MWe. 4 refs

Full MOX core design in ABWR

International Nuclear Information System (INIS)

Ihara, Toshiteru; Mochida, Takaaki; Izutsu, Sadayuki; Fujimaki, Shingo

2003-01-01

Electric Power Development Co., Ltd. (EPDC) has been investigating an ABWR plant for construction at Oma-machi in Aomori Prefecture. The reactor, termed FULL MOX-ABWR will have its reactor core eventually loaded entirely with mixed-oxide (MOX) fuel. Extended use of MOX fuel in the plant is expected to play important roles in the country's nuclear fuel recycling policy. MOX fuel bundles will initially be loaded only to less than one-third of the reactor, but will be increased to cover its entire core eventually. The number of MOX fuel bundles in the core thus varies anywhere from 0 to 264 for the initial cycle and, 0 to 872 for equilibrium cycles. The safety design of the FULL MOX-ABWR briefly stated next considers any probable MOX loading combinations out of such MOX bundle usage scheme, starting from full UO 2 to full MOX cores. (author)

QoS and QoE Aware N-Screen Multicast Service

Directory of Open Access Journals (Sweden)

Ghulam Sarwar

2016-01-01

Full Text Available The paper focuses on ensuring the quality-of-service (QoS and quality-of-experience (QoE requirements of users having heterogeneous devices in a multicast session. QoS parameters such as bit rate, delays, and packet losses are good indicators for optimizing network services but fall short in characterizing user perception (QoE. In N-Screen service, the users have different devices with heterogeneous attributes like screen size, resolution, and access network interface, and the users have different QoE on N-Screen devices with the same QoS parameters. We formulate the objective function of the N-Screen multicast grouping to ensure the minimum user’s QoE with smaller bandwidth requirement. We propose a dynamic user reassignment scheme to maintain and satisfy the QoE by adapting the user’s membership to the varying network conditions. The proposed schemes combine the available bandwidth and multimedia visual quality to ensure the QoS and QoE. In the network architecture, we introduce the functions of the QoS and QoE aware multicast group management and the estimation schemes for the QoS and QoE parameters. The simulation results show that the proposed multicast service ensures the network QoS and guarantees the QoE of users in the varying network conditions.

Compact reversed-field pinch reactors (CRFPR)

International Nuclear Information System (INIS)

Krakowski, R.A.; Hagenson, R.L.; Schnurr, N.M.; Copenhaver, C.; Bathke, C.G.; Miller, R.L.; Embrechts, M.J.

1986-01-01

The unique confinement properties of the poloidal-field-dominated Reversed-Field Pinch (RFP) are exploited to examine physics and technical issues related to a compact high-power-density fusion reactor. This resistive-coil, steady-state, toroidal device would use a dual-media (i.e., two separate coolants) power cycle that would be driven by a fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, and coils) having a power density and mass approaching pressurized-water-fission reactor values. A 1000-MWe(net) base case is selected from a comprehensive trade-off study to examine technological issues related to operating a high-power-density FPC. A general rationale outlining the need for improved fusion concepts is given, followed by a description of the RFP principle, a detailed systems and trade-off analysis, and a conceptual FPC design for the ∝ 20-MW/m 2 (neutrons) compact RFP reactor, CRFPR(20). Key FPC components are quantified, and full power-balance, thermal, and mechanical FPC integrations are given. (orig.)

Heterogeneous reactors

International Nuclear Information System (INIS)

Moura Neto, C. de; Nair, R.P.K.

1979-08-01

The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

Skin Inqjuries Reduce Survival and Modulate Corticosterone, C-Reactive Protein, Complement Component 3, IgM, and Prostaglandin E2 after Whole-Body Reactor-Produced Mixed Field (n + γ-Photons Irradiation

Directory of Open Access Journals (Sweden)

Juliann G. Kiang

2013-01-01

Full Text Available Skin injuries such as wounds or burns following whole-body γ-irradiation (radiation combined injury (RCI increase mortality more than whole-body γ-irradiation alone. Wound-induced decreases in survival after irradiation are triggered by sustained activation of inducible nitric oxide synthase pathways, persistent alteration of cytokine homeostasis, and increased susceptibility to systemic bacterial infection. Among these factors, radiation-induced increases in interleukin-6 (IL-6 concentrations in serum were amplified by skin wound trauma. Herein, the IL-6-induced stress proteins including C-reactive protein (CRP, complement 3 (C3, immunoglobulin M (IgM, and prostaglandin E2 (PGE2 were evaluated after skin injuries given following a mixed radiation environment that might be found after a nuclear incident. In this report, mice received 3 Gy of reactor-produced mixed field (n+γ-photons radiations at 0.38 Gy/min followed by nonlethal skin wounding or burning. Both wounds and burns reduced survival and increased CRP, C3, and PGE2 in serum after radiation. Decreased IgM production along with an early rise in corticosterone followed by a subsequent decrease was noted for each RCI situation. These results suggest that RCI-induced alterations of corticosterone, CRP, C3, IgM, and PGE2 cause homeostatic imbalance and may contribute to reduced survival. Agents inhibiting these responses may prove to be therapeutic for RCI and improve related survival.

Production reactor productivity improvement plan

International Nuclear Information System (INIS)

Leitz, E.E.

1980-01-01

The N Reactor complex, which is operated by UNC for DOE, is a unique facility and as such it is difficult to transfer technological developments and management innovations directly to the N Reactor operations. Therefore the approach to implementing an effective program was to start with the general systems philosophy and then progress into using those specific analytical and management techniques applicable to the unique situation (technologically and administratively) which existed at the N Reactor plant

TRATAMIENTO DE AGUAS RESIDUALES MEDIANTE REACTORES ANAERÓBICOS DE PLACAS VERTICALES PARALELAS EN ACRÍLICO TRATAMENTO DE ÁGUAS RESIDUÁRIAS POR REATORES ANAERÓBIOS DE PLACAS VERTICAIS PARALELAS EM ACRÍLICO WASTEWATER TREATMENT BY ANAEROBIC REACTORS OF VERTICAL PARALLEL PLATES IN ACRYLIC

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Guillermo Chaux F

2011-12-01

Full Text Available Algunos filtros anaeróbicos con lecho de piedra construidos en el departamento del Cauca (Colombia, están presentando problemas de colmatación. Si se reemplaza la piedra por placas verticales paralelas, se elimina el problema de obstrucción. Este documento presenta el desarrollo y resultados de una investigación que evaluó a escala de laboratorio el potencial de los reactores anaeróbicos de placas verticales paralelas en acrílico para remover contaminantes (materia orgánica y sólidos suspendidos. El reactor anaeróbico de placas paralelas en acrílico se desempeñó como tratamiento secundario; se alimentó con agua residual efluente de un Tanque Imhoff con concentraciones medias de 156 ± 14 mg/L de DB05, 438 ± 32 mg/L de DQO y 98 ±22 mg/L de sólidos suspendidos totales. Las remociones de DQO y DB05 en el reactor sobrepasan el 50% y la remoción de sólidos suspendidos sobrepasó el 60% para tiempos de detención de 24 horas. La facilidad en la operación del reactor lo hace viable como tratamiento biológico anaeróbico de aguas residuales previamente decantadasAlguns filtros anaeróbios com recheio de pedras construída no departamento de Cauca (Colombia estão apresentando problemas de obstrução. Se a pedra é substituída por placas verticais paralelas, evita o problema da obstrução. Este artigo apresenta o desenvolvimento e os resultados e no estudo realizado em escala de laboratório que avaliaram o potencial de reatores anaeróbios de placas verticais paralelas em acrílico para remover os contaminantes (sólidos suspensos e matéria orgânica. 0 reator anaeróbio de placas paralelas de acrílico serviu como tratamento secundário; foi alimentado com água residuária do efluente de um tanque Imhoff com concentrações médias de 156 ± 14 mg/L DB05, 438 ± 32 mg/L de DQO e 98 ±22 mg/L de sólidos suspensos totais. A remoção de DQO e DB05 no reator são mais de 50% ea remoção de sólidos em suspensão superior a 60

n-Heptane cool flame chemistry: Unraveling intermediate species measured in a stirred reactor and motored engine

KAUST Repository

Wang, Zhandong

2017-10-03

This work identifies classes of cool flame intermediates from n-heptane low-temperature oxidation in a jet-stirred reactor (JSR) and a motored cooperative fuel research (CFR) engine. The sampled species from the JSR oxidation of a mixture of n-heptane/O2/Ar (0.01/0.11/0.88) were analyzed using a synchrotron vacuum ultraviolet radiation photoionization (SVUV-PI) time-of-flight molecular-beam mass spectrometer (MBMS) and an atmospheric pressure chemical ionization (APCI) Orbitrap mass spectrometer (OTMS). The OTMS was also used to analyze the sampled species from a CFR engine exhaust. Approximately 70 intermediates were detected by the SVUV-PI-MBMS, and their assigned molecular formulae are in good agreement with those detected by the APCI-OTMS, which has ultra-high mass resolving power and provides an accurate elemental C/H/O composition of the intermediate species. Furthermore, the results show that the species formed during the partial oxidation of n-heptane in the CFR engine are very similar to those produced in an ideal reactor, i.e., a JSR.The products can be classified by species with molecular formulae of C7H14Ox (x = 0–5), C7H12Ox (x = 0–4), C7H10Ox (x = 0–4), CnH2n (n = 2–6), CnH2n−2 (n = 4–6), CnH2n+2O (n = 1–4), CnH2nO (n = 1–6), CnH2n−2O (n = 2–6), CnH2n−4O (n = 4–6), CnH2n+2O2 (n = 0–4, 7), CnH2nO2 (n = 1–6), CnH2n−2O2 (n = 2–6), CnH2n−4O2 (n = 4–6), and CnH2nO3 (n = 3–6). The identified intermediate species include alkenes, dienes, aldehyde/keto compounds, olefinic aldehyde/keto compounds, diones, cyclic ethers, peroxides, acids, and alcohols/ethers. Reaction pathways forming these intermediates are proposed and discussed herein. These experimental results are important in the development of more accurate kinetic models for n-heptane and longer-chain alkanes.

Self-Cleaning Boudouard Reactor for Full Oxygen Recovery from Carbon Dioxide

Science.gov (United States)

Coutts, Janelle; Hintze, Paul E.; Muscatello, Anthony C.; Gibson, Tracy L.; Captain, James G.; Lunn, Griffin M.; Devor, Robert W.; Bauer, Brint; Parks, Steve

2016-01-01

Oxygen recovery from respiratory carbon dioxide is an important aspect of human spaceflight. Methods exist to sequester the carbon dioxide, but production of oxygen needs further development. The current International Space Station Carbon Dioxide Reduction System (CRS) uses the Sabatier reaction to produce water (and ultimately breathing air). Oxygen recovery is limited to 50 because half of the hydrogen used in the Sabatier reactor is lost as methane, which is vented overboard. The Bosch reaction, which converts carbon dioxide to oxygen and solid carbon is capable of recovering all the oxygen from carbon dioxide, and is the only real alternative to the Sabatier reaction. However, the last reaction in the cycle, the Boudouard reaction, produces solid carbon and the resulting carbon buildup will eventually foul the nickel or iron catalyst, reducing reactor life and increasing consumables. To minimize this fouling and increase efficiency, a number of self-cleaning catalyst designs have been created. This paper will describe recent results evaluating one of the designs.

Catalytic-Dielectric Barrier Discharge Plasma Reactor For Methane and Carbon Dioxide Conversion

Directory of Open Access Journals (Sweden)

Istadi Istadi

2007-10-01

Full Text Available A catalytic - DBD plasma reactor was designed and developed for co-generation of synthesis gas and C2+ hydrocarbons from methane. A hybrid Artificial Neural Network - Genetic Algorithm (ANN-GA was developed to model, simulate and optimize the reactor. Effects of CH4/CO2 feed ratio, total feed flow rate, discharge voltage and reactor wall temperature on the performance of catalytic DBD plasma reactor was explored. The Pareto optimal solutions and corresponding optimal operating parameters ranges based on multi-objectives can be suggested for catalytic DBD plasma reactor owing to two cases, i.e. simultaneous maximization of CH4 conversion and C2+ selectivity, and H2 selectivity and H2/CO ratio. It can be concluded that the hybrid catalytic DBD plasma reactor is potential for co-generation of synthesis gas and higher hydrocarbons from methane and carbon dioxide and showed better than the conventional fixed bed reactor with respect to CH4 conversion, C2+ yield and H2 selectivity for CO2 OCM process. © 2007 BCREC UNDIP. All rights reserved.[Presented at Symposium and Congress of MKICS 2007, 18-19 April 2007, Semarang, Indonesia][How to Cite: I. Istadi, N.A.S. Amin. (2007. Catalytic-Dielectric Barrier Discharge Plasma Reactor For Methane and Carbon Dioxide Conversion. Bulletin of Chemical Reaction Engineering and Catalysis, 2 (2-3: 37-44.  doi:10.9767/bcrec.2.2-3.8.37-44][How to Link/DOI: http://dx.doi.org/10.9767/bcrec.2.2-3.8.37-44 || or local: http://ejournal.undip.ac.id/index.php/bcrec/article/view/8][Cited by: Scopus 1 |

Software é desenvolvido, e não fabricado como geladeira e fogão - Gerenciamento é essencial

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Antonio Mendes Silva Filho

2012-10-01

Full Text Available Software não é fabricado como geladeira e fogão que são montados. Software é um produto desenvolvido, de modo sistemático, através de um conjunto de atividades bem definidas, seguindo a execução disciplinada do gerenciamento de projeto. Software requer um processo de desenvolvimento e não um processo de ‘fabricação’, implicando que ele exigirá o uso da engenharia de software. Nesse sentido, este artigo explora a importancia da estimativa de tamanho de projeto para o desenvolvimento de software.

La entonación de enunciados declarativos e interrogativos en chino mandarín hablado por taiwaneses

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Wei-Li Kao

2014-11-01

Full Text Available El presente trabajo de investigación tiene el propósito de describir la entonación de los enunciados declarativos e interrogativos absolutos del chino mandarín hablado por taiwaneses. Teniendo en cuenta este objetivo, se ha elaborado un corpus de habla espontánea y se ha analizado siguiendo el método Análisis Melódico del Habla. El corpus está basado en 88 enunciados emitidos por 44 informantes, que provienen de las emisiones de dos programas de televisión taiwaneses. Los resultados obtenidos muestran que los enunciados declarativos e interrogativos absolutos tienen tendencia a finalizar con un contorno descendente leve, y además, se ha observado que el tono de los fonemas se modifica según la entonación del grupo fónico.

vacío - presión e inmersión

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R. Machuca-Velasco

2006-01-01

Full Text Available El presente trabajo se realizó con la finalidad de conocer la capacidad de absorción de preservadores de las maderas de Pinus arizonica (pino blanco, Pinus engelmanni (pino real, Pinus patula (pino colorado, Ceiba pentandra (ceiba, Manilkara zapota (chicozapote, Spondias mombin (jobo, Quercus insignis (chicalaba, Quercus laurina (laurelillo y Quercus oleoides (tesmol, aplicando los procesos de impregnación vacío-presión e inmersión con sales CCA (cobre, cromo, arsénico tipo C, a 2.5 %; pentaclorofenol (PCP, a 5 %; y creosota a 50 % de concentración. Los niveles de absorción se analizaron con técnicas univariadas. De acuerdo a los niveles de absorción alcanzados con el método de vacío-presión, con las sales CCA (cromo, cobre, arsénico los pinos, la ceiba y el jobo resultaron fáciles de impregnar, con pentaclorofenol y creosota sólo los pinos resultaron fáciles de impregnar. En cuanto al método de inmersión, con sales CCA, el pino colorado, la ceiba y el tesmol fueron fáciles de impregnar; con pentaclorofenol y creosota, los encinos, el chicozapote y el jobo fueron difíciles de impregnar. Para los pinos, la ceiba y el jobo, la mayor absorción con sales CCA se alcanzó con el método de impregnación a vacío-presión; para los encinos y el chicozapote, la mayor absorción con sales CCA fue con la inmersión prolongada.

Freqüência de diabetes mellitus e hiperglicemia em mulheres chagásicas e não-chagásicas

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Santos Vitorino Modesto dos

1999-01-01

Full Text Available Estudo retrospectivo de 647 mulheres com idade340 anos, atendidas no Hospital-Escola da FMTM, Uberaba-MG. As três sorologias para a doença de Chagas foram negativas nas controles (n = 285 e positivas nas chagásicas (n = 362, que foram classificadas nas formas indeterminada (n = 125, megas (n = 58 e cardíaca (n = 179. Diabetes mellitus foi definido por duas glicemias em jejum3140mg/dl e hiperglicemia por glicemia em jejum > 110mg/dl. Os grupos foram comparados pelos testes do c2, análise de variância, "t" de Student, Kruskal-Wallis e Mann-Whitney, considerando-se significativo p < 0,05. chagásicas e controles estavam pareadas quanto à idade, o índice de massa corporal e a cor. Diabetes mellitus foi mais freqüente na forma cardíaca (15,1%, comparada com as controles (7,4%, megas (7,4% e assintomáticas (5,6%, o mesmo ocorrendo com a hiperglicemia (37,4%, 26,7%, 25,9% e 27,2%, respectivamente, achados que estão de acordo com possível desnervação parassimpática causada pelo Trypanosoma cruzi e conseqüente predomínio da atividade simpática.

Prokaryotic diversity and dynamics in a full-scale municipal solid waste anaerobic reactor from start-up to steady-state conditions.

Science.gov (United States)

Cardinali-Rezende, Juliana; Colturato, Luís F D B; Colturato, Thiago D B; Chartone-Souza, Edmar; Nascimento, Andréa M A; Sanz, José L

2012-09-01

The prokaryotic diversity of an anaerobic reactor for the treatment of municipal solid waste was investigated over the course of 2 years with the use of 16S rDNA-targeted molecular approaches. The fermentative Bacteroidetes and Firmicutes predominated, and Proteobacteria, Actinobacteria, Tenericutes and the candidate division WWE1 were also identified. Methane production was dominated by the hydrogenotrophic Methanomicrobiales (Methanoculleus sp.) and their syntrophic association with acetate-utilizing and propionate-oxidizing bacteria. qPCR demonstrated the predominance of the hydrogenotrophic over aceticlastic Methanosarcinaceae (Methanosarcina sp. and Methanimicrococcus sp.), and Methanosaetaceae (Methanosaeta sp.) were measured in low numbers in the reactor. According to the FISH and CARD-FISH analyses, Bacteria and Archaea accounted for 85% and 15% of the cells, respectively. Different cell counts for these domains were obtained by qPCR versus FISH analyses. The use of several molecular tools increases our knowledge of the prokaryotic community dynamics from start-up to steady-state conditions in a full-scale MSW reactor. Copyright © 2012 Elsevier Ltd. All rights reserved.

Characterization of a fast to thermal neutron spectrum converter on PROSPERO reactor

Energy Technology Data Exchange (ETDEWEB)

Jacquet, X.; Authier, N.; Casoli, P.; Combacon, S. [CEA, Valduc Center, 21120 Is sur Tille (France); Calzavarra, Y. [ILL, Institut Laue Langevin, 38000 Grenoble (France)

2009-07-01

The PROSPERO reactor is located at CEA Valduc Center in France. The reactor is composed of an internal core made of High Enriched Uranium metal alloy surrounded by a reflector of depleted uranium. The reactor is used as a fast neutron spectrum source and is operated in delayed critical state with a continuous and steady power for several hours, which can vary from 3 mW to 3 kW, which is the nominal power. The flux at nominal power varies from 5.10{sup +10} n.cm{sup -2}/s at the reflector surface to 10{sup +7} n.cm{sup -2}/s at 5 meters from reactor axis. It has been decided to build a neutron energy converter allowing the production of a neutron thermal spectrum. As the core produces fast neutrons spectrum, we built a hollow cubic box of 50 cm x 50 cm x 50 cm with 10-cm-thick polyethylene bricks and placed one meter away from central reactor axis to moderate as much as possible neutrons to lower energies (E<0.6 eV). Analysis of the moderated flux inside the converter was performed using different activation foils such as indium or gold. We have developed a model of the experiment in the Monte Carlo neutron transport code TRIPOLI-4. A non-analogous transport calculation scheme was necessary to reproduce properly the experimental activities. The results of the calculated activations are within 4% of the experimental measurements given with 10% uncertainty (2 sigma). We show that the converter realizes thermalization of 80 % of the PROSPERO reactor fast neutrons below the cadmium threshold of 0.6 eV. Epithermal neutrons represent 15% of the spectrum and only 5% are in the fast neutron range above 1 MeV. The total flux at the center of the converter is 1.4 10{sup +9} n.cm{sup -2}/s at 3000 W

Hidrogenación de p-nitrofenol mediante el uso de catalizadores de Ir, Ni e Ir-Ni soportados en TiO2

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Hugo Alfonso Rojas Sarmiento

2012-06-01

Full Text Available Los catalizadores de Ir/TiO2,Ni/TiO2 e Ir-Ni/TiO2 fueron obtenidos mediante impregnación húmeda, a una concentración de 1% en peso del metal. Los catalizadores resultantes se caracterizaron mediante análisis de difracción de rayos X (DRX, fisisorción con nitrógeno a 77K, quimisorción de hidrógeno y temperatura programada de reducción (TPR. Los sólidos sintetizados fueron empleados como catalizadores en la reacción de hidrogenación de pnitrofenol para la obtención de p-aminofenol, importante intermediario para la síntesis de diversos analgésicos y antipiréticos. Los ensayos catalíticos se llevaron a cabo en un reactor tipo Batch a 0,62 MPa, 363K y etanol como disolvente. El progreso de la reacción fuemonitoreado por cromatografía de gases. El catalizador Ir/TiO2 exhibió el mayor nivel de conversión de p-nitrofenol (95,6% en 9 horas de reacción, lo cual fue atribuido a la presencia de sitios activos originados por el iridio y al efecto SMSI (interacción fuerte metal soporte por parte del iridio y níquel.Todos los catalizadores exhibieron una selectividad hacia el p- aminofenol del 100%.

Hidrogenación de p-nitrofenol mediante el uso de catalizadores de Ir, Ni e Ir-Ni soportados en TiO2

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Hugo Alfonso Rojas Sarmiento

2013-02-01

Full Text Available Los catalizadores de Ir/TiO2,Ni/TiO2 e Ir-Ni/TiO2 fueron obtenidos mediante impregnación húmeda, a una concentración de 1% en peso del metal. Los catalizadores resultantes se caracterizaron mediante análisis de difracción de rayos X (DRX, fisisorción con nitrógeno a 77K, quimisorción de hidrógeno y temperatura programada de reducción (TPR. Los sólidos sintetizados fueron empleados como catalizadores en la reacción de hidrogenación de pnitrofenol para la obtención de p-aminofenol, importante intermediario para la síntesis de diversos analgésicos y antipiréticos. Los ensayos catalíticos se llevaron a cabo en un reactor tipo Batch a 0,62 MPa, 363K y etanol como disolvente. El progreso de la reacción fuemonitoreado por cromatografía de gases. El catalizador Ir/TiO2 exhibió el mayor nivel de conversión de p-nitrofenol (95,6% en 9 horas de reacción, lo cual fue atribuido a la presencia de sitios activos originados por el iridio y al efecto SMSI (interacción fuertemetal soporte por parte del iridio y níquel.Todos los catalizadores exhibieron una selectividad hacia el p- aminofenol del 100%.

Hidrogenación e interesterificación del aceite de castaña de Brasil (Bertholletia excelsa

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Polakiewicz, Bronislaw

2001-08-01

Full Text Available Brazil nut oil (ACB was hydrogenated in a 1L Parr reactor, with Ni as catalyst, at the following process conditions: 175ºC, 3 atm, 60 min (GH1, 150ºC, 1 atm, 30 min (GH2 and 125ºC, 1 atm, 30 min (GH3. Different proportions of blends with ACB and GH1 and GH2 were prepared. These mixtures were interesterified at laboratory scale (0.75% of sodium metoxide, 60 min, 60-65ºC. Linoleic selectivity (Sl was 3.87 (GH1, 17.46 (GH2 and 17.46 (GH3. Linolenic selectivity (Sln was 2.3 for every reaction. It was observed different results for starting and interesterificated blends for the physical properties, for these parameters and for the interesterified fats, were applied a multiple regression. Results showed that consistency and solid fat content (SFC were dependent on the hydrogenated fats. Significant interactions were, in general, for the interesterified blends of ACB/GH1 and GH1/GH2, only for the consistency and not for the other properties.El aceite de castaña de Brasil (ACB fue hidrogenado, en un reactor Parr de 1 L, catalizador a base de Ni, y bajo las siguientes condiciones de proceso: 175ºC, 3 atm, 60 min (GH1, 150ºC, 1 atm, 30 min (GH2 y 125ºC, 1 atm, 30 min (GH3. Con las grasas resultantes se prepararon mezclas en diferentes proporciones de ACB con GH1 y GH2. Estas fueron interesterificadas a escala de laboratorio con 0.75% de metóxido de sodio, 60 min, 60-65ºC. En la hidrogenación la selectividad linoleica (Sl fue 3.87 (GH1, 17.46 (GH2 y 8.45 (GH3 y la selectividad linolénica (Sln fue 2.3 para las tres reacciones. A los parámetros de las propiedades físicas de los productos interesterificados, se aplicó un modelo de regresión múltiple. Los resultados mostraron que la consistencia y el contenido en grasa sólida dependían de la grasa hidrogenada, e indicaron que las interacciones fueron en general, significativas para las mezclas interesterificadas de ACB/GH1 y GH1/GH2 en cuanto a la consistencia, pero no en las

Towards nuclear fusion reactors

International Nuclear Information System (INIS)

1993-11-01

The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)

A full-scale UASB reactor for treatment of pig and cattle slaughterhouse wastewater with a high oil and grease content

Directory of Open Access Journals (Sweden)

L. A. S. Miranda

2005-12-01

Full Text Available This paper discusses the performance of an 800m³ full-scale UASB reactor in treating meat-packing plant and slaughterhouse effluents containing high concentrations of oil and grease (O&G (413-645 mg/L, resulting in a COD/O&G ratio of 26-32%. Those macromolecules were considered responsible for the unbalance of the system resulting in a total washout of the biomass. The removal of O&G from the influent using a physicochemical system (coagulation-flocculation improved the physical characteristics of the anaerobic sludge, controlling the biomass washout. Reactor performance was significantly improved when the COD/O&G ratio influent was maintained in the 10%. The COD and O&G removal rates obtained after implantation of the physicochemical system were 70-92% and 27-58%, respectively. The specific methanogenic activity (SMA of the biomass shows towards a tendency stabilisation and adaptation to the substrate influent. Pretreatment of the influent allowed the maximum organic load to be increased (1.46 to 2.43 Kg COD/m³.d and improved the quality of the effluent.

Neutron capture cross section measurement of 238U at the n TOF CERN facility with C6D6 scintillation detectors in the energy region from 1 eV to 700 keV

CERN Document Server

Mingrone, F.

2017-01-01

The aim of this work is to provide a precise and accurate measurement of the 238U(n,g) reaction cross section in the energy region from 1 eV to 700 keV. This reaction is of fundamental importance for the design calculations of nuclear reactors, governing the behaviour of the reactor core. In particular, fast reactors, which are experiencing a growing interest for their ability to burn radioactive waste, operate in the high energy region of the neutron spectrum. In this energy region most recent evaluations disagree due to inconsistencies in the existing measurements of up to 15%. In addition, the assessment of nuclear data uncertainty performed for innovative reactor systems shows that the uncertainty in the radiative capture cross-section of 238U should be further reduced to 1-3% in the energy region from 20 eV to 25 keV. To this purpose, addressed by the Nuclear Energy Agency as a priority nuclear data need, complementary experiments, one at the GELINA and two at the n_TOF facility, were proposed and carrie...

Vigorexia e níveis de dependência de exercício em frequentadores de academias e fisiculturistas

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Patrícia Tatiana Soler

2013-10-01

Full Text Available INTRODUÇÃO: Atualmente, a aparência é sinônimo de sucesso, saúde e determinação. De modo que a sociedade moderna oprima os indivíduos a seguir padrões estereotipados de beleza. E como herança dessa sociedade capitalista e egoísta surgem os transtornos psíquicos da aparência e as dependências psíquicas a eles associadas. OBJETIVOS: i comparar os níveis de vigorexia e de dependência ao exercício entre frequentadores de academias e fisiculturistas; ii relacionar as variáveis de prática de exercício físico (tempo de prática, frequência semanal e duração por sessão com as dimensões de frequência de vigorexia e de dependência ao exercício; e, iii comparar os níveis de vigorexia segundo grupos de dependência ao exercício (dependentes ou em risco, não dependente sintomático e não dependente assintomático. MÉTODOS: A amostra foi constituída por 151 frequentadores de academia (27,66 ± 6,54 anos e 27,56 ± 5,03 de IMC e 25 fisiculturistas (30,80 ± 5,54 anos e 26,72 ± 4,24 de IMC, todos do sexo masculino. Os participantes responderam à Escala de Dependência ao Exercício e ao Inventário de Dismorfia Muscular. A análise estatística envolveu procedimentos de análise descritiva, normalidade univariada, comparativa e correlacional. RESULTADOS: Os principais resultados evidenciaram: i não existir diferenças entre frequentadores de academias e fisiculturistas quanto aos níveis de vigorexia e de dependência ao exercício; ii que a duração da sessão de treino se correlaciona positivamente com a maioria das dimensões da dependência ao exercício; e, iii que o grupo classificado como dependente ou em risco revela níveis médios superiores de vigorexia. CONCLUSÃO: Por fim, constatou-se que tanto nos fisiculturistas como nos frequentadores de academias, quanto maior o nível de vigorexia, maior o nível de dependência ao exercício, sendo essa correlação maior em fisiculturistas.

DIREITOS HUMANOS E DIREITOS FUNDAMENTAIS: CONVERGÊNCIAS ENTRE JOAQUÍN HERRERA FLORES E LUIGI FERRAJOLI

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Leilane Serratine Grubba, UFSC, Brasil

2012-03-01

Full Text Available Resumo: O trabalho tem como objeto a reflexão sobre o fundamento dos Direitos Humanos e sua indissociável relação com os Direitos Fundamentais, no âmbito do Estado de Direito contemporâneo. Para tanto, procura estabelecer possíveis convergências teóricas entre Joaquín Herrera Flores e Luigi Ferrajoli. A análise dos Direitos Fundamentais, a partir da teoria garantista; e dos Direitos Humanos, com base da teoria crítica da reinvenção, é tanto um desafio teórico quanto prático do século XXI. Isso, em virtude da necessidade da investigação de seus fundamentos teóricos e possibilidades de implementação na dialética da práxis da vida em sociedade, para que as conquistas históricas de Direitos não sejam reduzidas à mera retórica hegemônico-conservadora de uma ordem global fundada na desigualdade e exploração. Percebe-se, então, que o constitucionalismo e os Direitos não importam em conquistas, mas em programas normativos a serem buscados diariamente, tal como a democracia e a paz. Nesse sentido, situa-se a importância do estudo dos direitos, estejam eles em normativas nacionais ou internacionais, bem como situá-los em sua transitoriedade e constante construção, em suma, em sua historicidade, para serem efetivamente garantidos. Palavras-chave: Garantismo. Teoria crítica. Direitos Humanos. Direitos Fundamentais.

Comparison of Simultaneous Nitrification and Denitrification for Three Different Reactors

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W. Khanitchaidecha

2015-01-01

Full Text Available Discharge of high NH4-N containing wastewater into water bodies has become a critical and serious issue due to its negative impact on water and environmental quality. In this research, the performance of three different reactors was assessed and compared with regard to the removal of NH4-N from wastewater. The highest nitrogen removal efficiency of 98.3% was found when the entrapped sludge reactor (ESR, in which the sludge was entrapped in polyethylene glycol polymer, was used. Under intermittent aeration, nitrification and denitrification occurred simultaneously in the aerobic and anaerobic periods. Moreover, internal carbon was consumed efficiently for denitrification. On the other hand, internal carbon consumption was not found to occur in the suspended sludge reactor (SSR and the mixed sludge reactor (MSR and this resulted in nitrogen removal efficiencies of SSR and MSR being 64.7 and 45.1%, respectively. Nitrification and denitrification were the main nitrogen removal processes in the aerobic and anaerobic periods, respectively. However, due to the absence of sufficient organic carbon, denitrification was uncompleted resulting in high NO3-N contents in the effluent.

PIK reactor construction status

International Nuclear Information System (INIS)

Konoplev, K.A.; Smolsky, S.L.

2001-01-01

The 100MW reactor PIK for fundamental researches has a thermal neutron flux of more than 10 15 n/cm 2 sec. This presentation outlines the construction state as of 2001, its prospects and completion tactics in the conditions of unstable finance. Construction of the reactor started in 1976. In 1986 construction of the building was completed and significant part of the installation work fulfilled. Construction of cooling systems was finished, the control panel assembled, and adjustment of the pump and gate valve control circuits started. After Chernobyl catastrophe, the USSR nuclear reactor safety requirements were revised. The PIK design did not meet these requirements and underwent considerable revision. The reconstruction design resulted in double the initial cost. Creation of the containment was the bulkiest part of the reconstruction. It brought about the need to disassemble the roofing of the building, dismantle all the equipment of the two upper floors, and lay up the equipment of the lower floors. As of 2001, construction in accordance with the revised design is at the stage of assemblage of the most important units, i.e. reactor itself, cooling system, heavy water system, and a number of auxiliary systems, such as depleted fuel storage, emergency cooling system etc. (orig.)

América: Identidad, Integración e Independencia

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Álvaro Acevedo Gutiérrez

2013-01-01

Full Text Available El nombre que se le ha dado al continente a trav s de los cinco siglos de miscegenaci n hist rica, ha marcado los niveles de dependencia e integraci n. En los siglos de dominaci n de Espa a se enfatiz en llamarlo Nuevo Orbe y Nuevo Mundo para diferenciarlo del viejo. Eran ante todo nombres impuestos. En la misma sincron a hist rica se le otorg el nombre de Indias , que marcaba una profunda discriminaci n tanto del continente como de todos sus habitantes, incluyendo a los mismos espa oles que al regresar a su pen nsula natal se les denominaba de manera peyorativa como Indianos . El nombre de Am rica se origin en una equivocaci n, como tambi n en una profunda injusticia con el Almirante Col n, sin embargo, los hombres de la independencia lo reivindicar n y con orgullo sus habitantes se autodenominar n como americanos o criollos americanos , d ndole sentido al origen de una identidad hist rica de Patria Grande . En tal sentido, el ideario de Sim n Bol var buscaba armonizar la identidad de Am rica con el proceso de independencia y a su vez a?rmar el proyecto de integraci n continental. El sue o de Bol var y de los americanos de principios del siglo XIX, estaba dirigido a sintetizar esos procesos en tres palabras claves: identidad, independencia e integraci n.

Potential applications of NbN composites in fusion reactor magnets

International Nuclear Information System (INIS)

Capone, D.W. II; Gray, K.E.; Kampwirth, R.T.; Ho, H.L.

1986-02-01

Recent projected requirements for large scale fusion reactor magnets call for the development of advanced superconducting materials capable of producing peak magnetic fields in excess of 15 T with current densities in the windings in excess of 2 x 10 3 A/cm 2 . These materials will be exposed to large stresses (up to 500 MPa) and neutron fluences as high as 10 22 n/cm 2 over the lifetime of the conductor. The demonstrated strain and radiation tolerance of NbN together with excellent superconducting properties make it a promising candidate to be used in a superconducting composite capable of satisfying these requirements. Our program at Argonne is directed towards demonstrating a method of fabrication which is capable of achieving these goals. Tests will be conducted on moderate lengths of NbN superconducting composites to verify the ability to achieve large overall current densities in magnetic fields up to 20 T. High field applications of NbN are also being investigated by groups in Japan and Germany

Imagem corporal em mulheres adultas vs. meia-idade e idosas praticantes e não praticantes de hidroginástica

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Simone Valéria Dias Souto

2016-06-01

Full Text Available O objetivo do presente estudo foi comparar a imagem corporal de mulheres adultas vs. meia-idade e idosas praticantes e não praticantes de hidroginástica. A amostra foi constituída por 300 mulheres brasileiras com idades compreendidas entre 20 e 83 anos (48.96 ± 15.41, divididas em quatro grupos: Grupo 1 = 75 mulheres adultas não praticantes com idades entre 20 e 49 anos; Grupo 2 = 75 mulheres de meia-idade e idosas não praticantes com idades entre 50 e 82 anos; Grupo 3 = 75 mulheres de meia-idade e idosas praticantes com idade entre 50 e 83 anos e Grupo 4 = 75 mulheres adultas praticantes com idade entre 20 e 49 anos. O instrumento utilizado para determinar à imagem corporal actual e desejada foi a Escala de Desenhos de Silhuetas descrita por Stunkard. Os resultados demonstram que não houve diferenças significativas com a insatisfação da imagem corporal entre os grupos de mulheres adultas vs. meia-idade e idosas vs. praticantes vs. não praticantes (p>0.05. Entretanto, houve diferenças significativas entre a insatisfação com a imagem corporal actual vs. ideal para todos os grupos do estudo (p<0.05. Conclui-se que a idade e a prática da hidroginástica não são factores determinantes na percepção que as mulheres têm da imagem corporal.

Nitrogen Removal from Milking Center Wastewater via Simultaneous Nitrification and Denitrification Using a Biofilm Filtration Reactor

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Seung-Gun Won

2015-06-01

Full Text Available Milking center wastewater (MCW has a relatively low ratio of carbon to nitrogen (C/N ratio, which should be separately managed from livestock manure due to the negative impacts of manure nutrients and harmful effects on down-stream in the livestock manure process with respect to the microbial growth. Simultaneous nitrification and denitrification (SND is linked to inhibition of the second nitrification and reduces around 40% of the carbonaceous energy available for denitrification. Thus, this study was conducted to find the optimal operational conditions for the treatment of MCW using an attached-growth biofilm reactor; i.e., nitrogen loading rate (NLR of 0.14, 0.28, 0.43, and 0.58 kg m−3 d−1 and aeration rate of 0.06, 0.12, and 0.24 m3 h−1 were evaluated and the comparison of air-diffuser position between one-third and bottom of the reactor was conducted. Four sand packed-bed reactors with the effective volume of 2.5 L were prepared and initially an air-diffuser was placed at one third from the bottom of the reactor. After the adaptation period of 2 weeks, SND was observed at all four reactors and the optimal NLR of 0.45 kg m−3 d−1 was found as a threshold value to obtain higher nitrogen removal efficiency. Dissolved oxygen (DO as one of key operational conditions was measured during the experiment and the reactor with an aeration rate of 0.12 m3 h−1 showed the best performance of NH4-N removal and the higher total nitrogen removal efficiency through SND with appropriate DO level of ~0.5 mg DO L−1. The air-diffuser position at one third from the bottom of the reactor resulted in better nitrogen removal than at the bottom position. Consequently, nitrogen in MCW with a low C/N ratio of 2.15 was successfully removed without the addition of external carbon sources.

Co-creación e innovación abierta: Revisión sistemática de literatura

Directory of Open Access Journals (Sweden)

2018-01-01

Full Text Available La ciencia abierta, como bien común, abre posibilidades para el desarrollo de las naciones a través de innovaciones y construcciones colaborativas que ayudan a democratizar el conocimiento. Los avances en la materia aún son incipientes y el triángulo ciencia abierta, cocreación del conocimiento e innovación abierta se presenta como una oportunidad de generar un aporte original, desde la investigación, para la teoría y las prácticas educativas abiertas. En el estudio se analizaron los artículos que abordan este triángulo, con el fin de identificar los contextos y retos que se presentan en la innovación y en la cocreación de conocimiento para impulsar la ciencia abierta. El método fue una revisión sistemática de literatura (SLR de 168 artículos publicados en acceso abierto, de enero 2014 a mayo 2017, en las bases de datos Web of Science y Scopus. La validación se dio con los criterios de la Universidad de York: inclusión y exclusión, pertinencia, evaluación de calidad / validez de los estudios y descripción de datos. Los hallazgos reflejan que los contextos de mayor publicación sobre el tema son los de Estados Unidos y Brasil, en los sectores empresariales y académicos (seguido de cerca por el sector social y los retos se abren en las posibilidades de innovación, apertura e investigación. Se concluye que el contexto y las prácticas de colaboración son elementos sustanciales para la innovación y la ciencia abierta.

Method of avoiding hazards resulting from accidents in water-cooled nuclear reactors

International Nuclear Information System (INIS)

Dorner, S.; Schretzmann, K.; Schumacher, G.

1984-01-01

In water-cooled reactors, e.g. BWRs and PWRs, elemental hydrogen is released by hydrolysis (in-leakage). In case of an accident in these reactors or at emergency cooling of e.g., a gas-cooled reactor with water additional hydrogen is produced by chemical reactions of the water with the cladding material. In order to prevent hydrogen pressurizing and the formation of a detonating gas mixture, dry powder containers are provided for in the endangered compartments of the reactor. In case of danger powdered CuO, MnO 2 , Fe 2 O 3 , or CdO, the oxygen content of which recombines with the hydrogen, is ejected from them. In addition, an extinguishing substance with an anticatalytic resp. inhibition effect and/or an inert gas of the group N 2 , He, Ar, CO 2 may be admixed to the powder resp. powder mixture. (orig./PW)

A report of the overall working group of the AEC Committee on Development of Advanced Power Reactors

International Nuclear Information System (INIS)

1979-01-01

The AEC Committee on Development of Advanced Power Reactors was set up in April, 1978, following on the previous AEC Special Committee on Development of Advanced Power Reactors, in order to study on the complementary power reactors between current LWRs and future FBRs. The subjects of study by the overall working group are the status of advanced power reactors in views of the nuclear fuel cycle, the impacts on industries, the selection of reactor types under present international circumstances, and the evaluation of advanced power reactors in their technology and economy. The following matters are described: evaluations in view of the nuclear fuel cycle, i.e. the features of the ATR of Japan and CANDU reactors of Canada; international problems concerning nuclear nonproliferation and securing of uranium; problems in the diversification of power reactor types concerning the expenditure and technology; problems of technology in the ATR of Japan, CANDU reactors of Canada and Pu utilization for LWRs; and the economy of D 2 O power reactors, i.e. the ATR of Japan and CANDU reactors of Canada. (J.P.N.)

Comprensión lectora y resolución de problemas en estudiantes de Educación Primaria Bilingüe en comunidades shipibas

Directory of Open Access Journals (Sweden)

Dulio Oseda Gago

2014-12-01

Full Text Available Objetivos: Determinar la relación entre la comprensión lectora y la resolución de problemas en estudiantes del sexto grado de educación primaria bilingüe de las comunidades shipibas del distrito de Yarinacocha, Ucayali en el 2014. Métodos: Investigación de enfoque cuantitativo, tipo básica, diseño descriptivo - correlacional. La muestra fue tomada probabilísticamente conformada por 56 estudiantes del sexto grado de Educación Primaria Bilingüe en las comunidades shipibas del distrito de Yarinacocha el cual se determinó con un nivel de significancia del 5 %. Según el diseño, se utilizó los estadígrafos de la estadística descriptiva e inferencial, para contrastar la hipótesis se hizo uso de la prueba r de Pearson y la t de Student, que nos permitió deducir que existe una correlación directa y significativa (r=0,592. Resultados: La significancia obtenida (0,003 resultó inferior al nivel de significación propuesta (α = 0,05 por lo que se decide rechazar la hipótesis nula, afirmando que en términos generales, existe relación directa y significativa entre la comprensión lectora y la resolución de problemas matemáticos en los estudiantes del sexto grado de Educación Primaria Bilingüe evaluados en las comunidades Shipibas del distrito de Yarinacocha. Conclusiones: Se afirma que existe una relación directa entre la comprensión lectora y la resolución de problemas en estudiantes del sexto grado de Educación Primaria Bilingüe en las comunidades shipibas del distrito de Yarinacocha, 2014.

Nuclear reactors

International Nuclear Information System (INIS)

Middleton, J.E.

1977-01-01

Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

Producción de hidrógeno a partir del tratamiento anaerobio de vinazas en un reactor UASB

Directory of Open Access Journals (Sweden)

César González-Ugalde

2014-09-01

Bajo condiciones mesofílicas (37 °C, un pH de operación de aproximadamente 5,50, una concentración del sustrato de 20 000 mg DQO/L y un tiempo de retención hidráulica (TRH de seis horas, la producción promedio de hidrógeno obtenida en el reactor UASB fue de 1,68 mL H2/h/L, con una tasa máxima de 13,4 mL H2/h/L. El porcentaje de remoción de DQO en el proceso de fermentación alcanzó valores máximos del 43%, con un promedio cercano al 20%. Tanto la producción de hidrógeno como la remoción de DQO presentaron una dependencia inversamente proporcional al TRH. Los resultados obtenidos en este estudio demuestran que la fermentación anaerobia en un reactor UASB abre la posibilidad de utilizar las vinazas para producir hidrógeno molecular de forma sostenible.

The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Younggwang nuclear power plant unit1

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae [Korea Atomic Energy Research Institute, Taejeon (Korea)

2001-08-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 1 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.555E+18, 1.662E+19, 3.358E+19, and 4.521E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.859 for the 1st through 4th testing and the calculational uncertainty, 11.80% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.551E+19 n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.929E+19, 4.880E+19, 5.831E+19 and 6.782E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 1 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 4 refs., 41 figs., 35 tabs. (Author)

The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Yonggwang nuclear power plant unit 2

Energy Technology Data Exchange (ETDEWEB)

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

2001-02-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.762E+18, 1.5391E+19, 3.5119E+19, and 4.2610E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.899 for the 1st through 4th testing and the calculational uncertainty, 12.3% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.357E+19 n/cm{sup 2} based on the end of 11th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.525E+19, 4.337E+19, 5.148E+19 and 5.960E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 42 tabs. (Author)

Monte Carlo simulation of irradiation of MTR fuel plates in the BR2 reactor using a full-scale 3-d model with inclined channels

International Nuclear Information System (INIS)

Kuzminov, V. V; Koonen, E.; Ponsard, B.

2002-01-01

A three-dimensional full-scale Monte Carlo model of the BR2 reactor has been developed for simulation of irradiation conditions of materials and fuel loaded in various irradiation devices. This new reactor model includes a detailed geometrical description of the inclined reactor channels, the irradiation devices loaded in these channels including the materials to be tested/loaded in these devices, the burn-up of the BR2 fuel elements and the poisoning of the beryllium matrix. Recently a benchmark irradiation of new irradiation device for testing and qualification of MTR fuel plates has been performed. For this purpose the detailed irradiation conditions of fuel plates had to be predetermined. Monte Carlo calculations of neutron fluxes and heat load distributions in irradiated MTR fuel plates were performed taking into account the contents of all loaded experimental devices in the reactor channels. A comparison of the calculated and measured values of neutron fluxes and of heat loads in the BR2 reactor is presented in this paper. The comparison is part of the validation process of the new reactor model. It also serves to establish the capability to conduct a fuel plate irradiation program under requested and well- known irradiation conditions. (author)

Implementación de la Estrategia y Plan de Acción de eSalud en la República Argentina, 2011 - 2013

Directory of Open Access Journals (Sweden)

Myrna Marti

2014-06-01

Full Text Available Como resultado de la aprobación de la Estrategia y Plan de Acción Regional de eSalud de la Organización Panamericana de la Salud (OPS/Organización Mundial de la Salud (OMS en 2011, en la República Argentina se desarrolló un proyecto de cooperación técnica entre la Representación de la OPS, la OMS y el Ministerio de Salud de la Nación que tiene como objetivo incorporar la estrategia a nivel nacional. Para ello, se desarrollaron una serie de reuniones, documentos y actividades entre las cuales se destacan la realización de un mapeo de iniciativas de eSalud en el país que incorporan tecnologías de la información y la comunicación (TIC en salud, identificación y selección de experiencias exitosas en eSalud, fomento de la cooperación horizontal, programa de colaboración virtual, programa de alfabetización digital y taller de eSalud. Los resultados obtenidos en este proceso se relacionan con el cumplimiento de los cuatro objetivos estratégicos de la Estrategia y Plan de Acción de eSalud de la OPS/OMS.

Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

Energy Technology Data Exchange (ETDEWEB)

Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

2013-07-01

A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

International Nuclear Information System (INIS)

Armstrong, J.; Hamilton, H.; Hyland, B.

2013-01-01

A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

Final report for the 5th surveillance test of the reactor pressure vessel material (capsule Y) of Yonggwang Nuclear Power Plant unit 2

Energy Technology Data Exchange (ETDEWEB)

Lee, Sam Lai; Kim, ByoungChul; Chang, Kee Ok (and others)

2006-02-15

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X, W and Y are 5.777E+18, 1.5371E+19, 3.7634E+19, 4.3045E+19, and 4.8662E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.953 for the 1st through 5th testing and the calculational uncertainty,7.2% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.659E+19n/cm{sup 2} based on the end of 13th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 48, 56 and 64EFPY would reach 3.625E+19, 5.293E+19, 6.127E+19 and 6.960E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the Pressurized Thermal Shock(PTS) during the operation until design life.

Crescimento e marcha de absorção de nutrientes da melancieira fertirrigada com diferentes doses de N e K

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Fabiola Pascoal de Nogueira

2014-07-01

Full Text Available 800x600 Na cultura da melancia, a nutrição mineral é um dos fatores mais importante que contribui diretamente na produtividade e qualidade dos frutos. O nitrogênio e o potássio são os dois nutrientes mais exigidos, e deve ser aplicado de acordo com as exigências de cada cultivar.O trabalho teve como objetivo avaliar o crescimento e a marcha de absorção de cultivares de melancia sob diferentes doses de nitrogênio e potássio. O experimento, com as duas variedades de melancia, foi conduzido no delineamento experimental em blocos casualizados no esquema de parcelas subdivididas 13 x 2, sendo o primeiro fator representado pela combinação de doses de nitrogênio (N e potássio (K em arranjo definido segundo o modelo: 2 x 2k+ 2k + 1, sendo k o número de fatores estudados (N e K.  Os totais de N, P e K acumulados no tecido vegetal não variaram com as doses de N ou K aplicadas em fertirrigação.Os máximos valores acumulados de N total foram aos 62 (10,9 g planta-1 e 61 DAT (13,31 g planta-1, para as cultivares Quetzali e Leopard. Para P total o valor máximo ocorreu aos 57 DAT (1,88 g planta-1 e para o K aos 64 DAT (15,15 g planta-1, independentes da cultivar. Normal 0 21 false false false PT-BR X-NONE X-NONE

Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

2016-05-15

KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

Bare nucleus S(E) factor of the 2H(d,p)3H and 2H(d,n)3He reactions via the Trojan Horse Method

International Nuclear Information System (INIS)

Tumino, A; Spitaleri, C; Kiss, G G; Cognata, M La; Lamia, L; Pizzone, R G; Rapisarda, G G; Romano, S; Sergi, M L; Spartà, R; Mukhamedzhanov, A M; Typel, S; Aliotta, M; Burjan, V; Kroha, V; Hons, Z; Mrazek, J; Piskor, S; Santo, M Gimenez del

2012-01-01

The Trojan Horse Method was applied for the first time to the 2 H(d,p) 3 H and 2 H(d,n) 3 He reactions by measuring the 2 H( 3 He,p 3 H) 1 H and 2 H( 3 He,n 3 He) 1 H processes in quasi free kinematics. The 3 He+d experiment was performed at 18 MeV, corresponding the a d-d energy range from 1.5 MeV down to 2 keV. This range overlaps with the relevant region for Standard Big Bang Nucleosynthesis as well as with the thermal energies of future fusion reactors and deuterium burning in the Pre Main Sequence phase of stellar evolution. This is the first pioneering experiment in quasi free regime where the charged spectator is detected. Both the energy dependence and the absolute value of the bare nucleus S(E) factors have been extracted for the first time. They deviate by more than 15% from available direct data with new S(0) values of 57.4±1.8 MeVb for 3 H+p and 60.1±1.9 MeVb for 3 He+n. None of the existing fitting curves is able to provide the correct slope of the new data in the full range, thus calling for a revision of the theoretical description. This has consequences in the calculation of the reaction rates with more than a 25% increase at the temperatures of future fusion reactors.

Analysis of influence of fast neutron fluence irradiated to Beryllium element of The RSG-GAS reactor

International Nuclear Information System (INIS)

Sri Kuntjoro

2010-01-01

Analysis of influence fast neutron fluence irradiated to the RSG-GAS beryllium reflector have been done. Methods of analysis was carried out by measuring fluxes neutron in beryllium element and block position that function as reflector.The calculation done for determination it is there any influence of neutron as long as beryllium in the core. Besides that, visualization done to make sure it there is any deformation at beryllium as effect of irradiation. Fluxes and fluences of beryllium element measurement result in 200 kW reactor power are 2.30E+07 n/cm 2 .sec and 4.19E+17 n/cm 2 in position E-2, 3.70E+07 n/cm 2 s and 6.74E+17 n/cm 2 in position J-8, 2.19E+12 n/cm 2 s and 3.99E+22 n/cm 2 in position. Measurement results in the position B-3 are 2.12E+12 n/cm 2 s and 3.86E+22 n/cm 2 in position G-10 respectively. Other result are fluxes and fluence in beryllium block, those are 5,02E+07 n/cm 2 s and 9,15E+17 n/cm 2 in position (5-6), and 2,32E+07 n/cm 2 s and 4,23E+17 n/cm 2 in position (C-D). Deformation (L/L) results for beryllium element are 1,12E-08 in position E-2, 1,84E-08 in position J-8, 1,60E-03 in position B-3, and 1,55E-03 in position G-10. In beryllium block deformation results are 2,52E-08 in position (5-6) and 1,13E-08 in position (C-D). Those results are shown unseen deformation in beryllium element and beryllium block and demonstrably by visual observation in reactor hot cell. (author)

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

Política e historia en la globalización altermundista

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Francisco Jorge Leira Castiñeira

2011-04-01

Full Text Available El artículo expone la configuración de las ideologías que protagonizan el enfrenamiento sociopolítico de los últimos años, pero que no permiten un punto de encuentro para la solución de los problemas globales. Me refiero al pensamiento único liberal y el pensamiento altermundista. Es ineludible fomentar un pensamiento plural y de debate para consensuar una solución en la erradicación de la desigualdad social y generar un ecosistema sostenible. Palabras clave: pensamiento único, globalización, altermudialización, neoconservador, liberalismo, antiglobalización.___________________________Abstract:In the last decades, the predominant models of thinking involved in the socio-political confrontation have been shaped: the liberal “pensée unique” and the alterworldist thinking. The liberal “pensée unique” has left no room for criticism, as it has strengthened its social and political influence. The alterworldist thinking arises in response to the acts of injustice in the current political system.This article focuses on their historical and political evolution.  It puts forward the promotion of pluralistic and debate-oriented thinking to help generate an equitable and ecologically sustainable social system.Keywords: “pensée unique”, globalization, alterworldist, neoconservadurism, liberalism, antiglobalization.

Nuclear reactor, reactor core thereof, and device for constituting the reactor

International Nuclear Information System (INIS)

Takiyama, Masashi.

1994-01-01

A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

Estudio de la recuperación de cromo hexavalente mediante un reactor electroquímico de compartimentos separados por separadores cerámicos

OpenAIRE

REYES PINEDA, HENRY

2011-01-01

La Tesis Doctoral "Estudio de la recuperación de cromo hexavalente mediante un reactor electroquímico de compartimentos separados por separadores cerámicos" se centra en la posibilidad de recuperación del cromo hexavalente procedente de las disoluciones de mordentado de las industrias de metalizado de plásticos mediante la utilización de un reactor electroquímico de compartimentos separados por separadores cerámicos fabricados a diferente presión y diferente composición de almidón. Con la rec...

Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

1.1 Need for Neutronics Calculations—An accurate calculation of the neutron fluence and fluence rate at several locations is essential for the analysis of integral dosimetry measurements and for predicting irradiation damage exposure parameter values in the pressure vessel. Exposure parameter values may be obtained directly from calculations or indirectly from calculations that are adjusted with dosimetry measurements; Guide E944 and Practice E853 define appropriate computational procedures. 1.2 Methodology—Neutronics calculations for application to reactor vessel surveillance encompass three essential areas: (1) validation of methods by comparison of calculations with dosimetry measurements in a benchmark experiment, (2) determination of the neutron source distribution in the reactor core, and (3) calculation of neutron fluence rate at the surveillance position and in the pressure vessel. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is th...

Evaluación del comportamiento hidrodinámico como herramienta para optimización de reactores anaerobios de crecimiento en medio fijo

OpenAIRE

Andrea Pérez; Patricia Torres

2008-01-01

Las condiciones de flujo no ideal en los reactores afectan su desempeño; las causas comunes son cortos circuitos, zonas muertas y recirculación interna por corrientes cinéticas y/o de densidad. En este estudio se optimizó el diseño de un filtro anaerobio a escala real que trata las aguas residuales del proceso de extracción de almidón de yuca, el cual presentaba problemas de represamiento y bajas eficiencias de remoción. La evaluación del comportamiento hidrodinámico inicial mostró la presenc...

Difusividade e condutividade hidráulica não saturada de substratos

Directory of Open Access Journals (Sweden)

Beatriz S. Conceição

2014-06-01

Full Text Available A quantificação dos processos hidráulicos que ocorrem dentro do recipiente durante e entre irrigações, é essencial para a gestão eficaz da irrigação e adubação em substratos. Os testes foram feitos no laboratório de hidráulica da Universidade Federal de Lavras com o objetivo de determinar a condutividade hidráulica não saturada de substratos usando-se o método proposto por Bruce & Klute (1956 em seis substratos (S1, S2, S3, S4, S5 e S6 com quatro repetições. Foram realizados ensaios de fluxo horizontal visando estimar a difusividade e, por seu meio, a condutividade hidráulica não saturada (K(θ. A taxa com que a umidade avança no espaço e no tempo ocorre de maneira diferenciada entre os substratos avaliados destacando-se os substratos S5 (casca de pinus, vermiculita e turfa e S2 (cinza, turfa e outros materiais, que tiveram menor e maior difusividade, respectivamente. Após o aumento da tensão ocorre uma nítida diminuição da K(θ, especialmente no substrato S3 (casca de pinus e terra vegetal enquanto para o substrato S5 os valores foram tão pequenos que não se destacam dentre os demais. A condutividade hidráulica não saturada aumenta com o aumento da umidade, de forma exponencial.

La difusión e interpretación del Patrimonio cultural e histórico-educativo como tema de estudio en el Centro de Formación Permanente de la Universidad de Sevilla

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Pablo Álvarez Domínguez

2010-06-01

Full Text Available (ES Recoge este texto el resultado de una experiencia educativa organizada por el Museo Pedagógico Andaluz y desarrollada el pasado mes de mayo de 2010 en el Centro de Formación Permanente de la Universidad de Sevilla, con la finalidad de propiciar una reflexión teórico-práctica ligada a las nuevas posibilidades didácticas y de difusión del patrimonio cultural e histórico-educativo que se pueden desplegar en el ámbito museístico, con el fin de facilitar su estudio, comprensión y valoración.

Nuclear reactor coolant channels

International Nuclear Information System (INIS)

Macbeth, R.V.

1978-01-01

A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

Investigación cualitativa e imaginación epidemiológica, una relación vital

Directory of Open Access Journals (Sweden)

Popay J.

2003-01-01

Full Text Available Este artículo presume que la «imaginación epidemiológica» tiene un papel relevante en el futuro desarrollo e implementación de políticas cuya finalidad sea mejorar la salud de la población y reducir las desigualdades de salud entre países y dentro de ellos. Sin embargo, si obviamos la contribución de la investigación cualitativa a la epidemiología fracasaremos en el desarrollo de ese potencial. Este artículo brevemente describe qué es la investigación cualitativa desde un punto de vista epidemiológico -qué tipo de «conocimiento» genera- y aborda cuestiones metodológicas (aproximaciones sobre recogida de datos, análisis e interpretación de resultados. Se presentan dos modelos diferentes sobre la relación entre la investigación cualitativa y cuantitativa. El primer modelo, denominado de intensificación, asume que los resultados de la investigación cualitativa enriquecen los resultados obtenidos mediante investigación cuantitativa y sugiere 3 roles para la investigación cualitativa: generar hipótesis que se podrán probar en la investigación cuantitativa, ayudar a construir medidas más sofisticadas de fenómenos sociales y explicar resultados sorprendentes obtenidos mediante una aproximación cuantitativa. Por el contrario, el modelo epistemológico sugiere que la investigación cualitativa es diferente de la cuantitativa en el sentido de que realiza una contribución única: estudia aspectos que otras aproximaciones no pueden analizar; mejora y amplía el conocimiento al profundizar en aspectos conceptuales y teóricos; cambia el equilibrio de poder entre los investigadores y el objeto de investigación, y por último desafía los métodos que utiliza la epidemiología tradicional para conocer el mundo social. Este artículo ilustra los diferentes tipos de contribuciones con ejemplos de la investigación cualitativa y finalmente discute cómo la investigación cualitativa "digna de confianza" puede evaluarse.

TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

Willingham, C.E.; Killgore, M.R.

1988-01-01

Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

Benthic megafauna at deep-sea study areas W-N and E-N

International Nuclear Information System (INIS)

Keller, C.H.

1985-01-01

Photographs were utilized determine what megafaunal organisms live at areas E-N and W-N and in what densities. Life history characteristics gleaned from the literature as well as from the photographs are presented to establish these organisms' roles in oceanic food chains. In area W-N, the most abundant megafaunal group is the echinoderms. These are the asteroids, crinoids, echinoids, holothuroids, and ophiuroids. Dominant amongst the echinoderms are the ophiuroids, which account for about 46% of the total density. Taxonomic and photographic resolution at area E-N are both poor. Numbers of megafaunal organisms cannot be directly compared between areas E-N and W-N. Differing photographic techniques severely limit comparisons between the two areas. Nevertheless, megafauna seen at area E-N is apparently less dense and probably less diverse than at area W-N. 30 references, 8 figures, 8 tables

Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor