Review of the C-nat(n,gamma) cross section and criticality calculations of the graphite moderated reactor BR1

A review of the experimental data for natC(n,c) and 12C(n,c) was made to identify the origin of the natC capture cross sections included in evaluated data libraries and to clarify differences observed in neutronic calculations for graphite moderated reactors using different libraries. The performance of the JEFF-3.1.2 and ENDF/B-VII.1 libraries was verified by comparing results of criticality calculations with experimental results obtained for the BR1 reactor. This reactor is an air-cooled reactor with graphite as moderator and is located at the Belgian Nuclear Research Centre SCK-CEN in Mol (Belgium). The results of this study confirm conclusions drawn from neutronic calculations of the High Temperature Engineering Test Reactor (HTTR) in Japan. Furthermore, both BR1 and HTTR calculations support the capture cross section of 12C at thermal energy which is recommended by Firestone and Révay. Additional criticality calculations were carried out in order to illustrate that the natC thermal capture cross section is important for systems with a large amount of graphite. The present study shows that only the evaluation carried out for JENDL-4.0 reflects the current status of the experimental data.

Methodologies for an improved prediction of the isotopic content in high burnup samples. Application to Vandellos-II reactor core

Fuel cycles are designed with the aim of obtaining the highest amount of energy possible. Since higher burnup values are reached, it is necessary to improve our disposal designs, traditionally based on the conservative assumption that they contain fresh fuel. The criticality calculations involved must consider burnup by making the most of the experimental and computational capabilities developed, respectively, to measure and predict the isotopic content of the spent nuclear fuel. These high burnup scenarios encourage a review of the computational tools to find out possible weaknesses in the nuclear data libraries, in the methodologies applied and their applicability range. Experimental measurements of the spent nuclear fuel provide the perfect framework to benchmark the most well-known and established codes, both in the industry and academic research activity. For the present paper, SCALE 6.0/TRITON and MONTEBURNS 2.0 have been chosen to follow the isotopic content of four samples irradiated in the Spanish Vandellós-II pressurized water reactor up to burnup values ranging from 40 GWd/MTU to 75 GWd/MTU. By comparison with the experimental data reported for these samples, we can probe the applicability of these codes to deal with high burnup problems. We have developed new computational tools within MONTENBURNS 2.0. They make possible to handle an irradiation history that includes geometrical and positional changes of the samples within the reactor core. This paper describes the irradiation scenario against which the mentioned codes and our capabilities are to be benchmarked.

Neutronics and activation of the preliminary reaction chamber of HiPER reactor based in a SCLL blanket

The HiPER reactor design is exploring different reaction chambers. In this study, we tackle the neutronicsand activation studies of a preliminary reaction chamber based in the following technologies: unpro-tected dry wall for the First Wall, self-cooled lead lithium blanket, and independent low activation steelVacuum Vessel. The most critical free parameter in this stage is the blanket thickness, as a function ofthe6Li enrichment. After a parametric study, we select for study both a ?thin? and ?thick? blanket, with?high? and ?low?6Li enrichment respectively, to reach a TBR = 1.1. To help to make a choice, we com-pute, for both blanket options, in addition to the TBR, the energy amplification factor, the tritium partialpressure, the203Hg and210Po total activity in the LiPb loop, and the Vacuum Vessel thickness requiredto guarantee the reweldability during its lifetime. The thin blanket shows a superior performance in thesafety related issues and structural viability, but it operates at higher6Li enrichment. It is selected forfurther improvements. The Vacuum Vessel shows to be unviable in both cases, with the thickness varyingbetween 39 and 52 cm. Further chamber modifications, such as the introduction of a neutron reflector,are required to exploit the benefits of the thin blanket with a reasonable Vacuum Vessel.

Nuclear Data Uncertainty Propagation to Reactivity Coefficients of a Sodium Fast Reactor

The assessment of the uncertainty levels on the design and safety parameters for the innovative European Sodium Fast Reactor (ESFR) is mandatory. Some of these relevant safety quantities are the Doppler and void reactivity coefficients, whose uncertainties are quantified. Besides, the nuclear reaction data where an improvement will certainly benefit the design accuracy are identified. This work has been performed with the SCALE 6.1 codes suite and its multigroups cross sections library based on ENDF/B-VII.0 evaluation.

Heat losses in a CVD reactor for polysilicon: comprehensive model and experimental validation

This work addresses heat losses in a CVD reactor for polysilicon production. Contributions to the energy consumption of the so-called Siemens process are evaluated, and a comprehensive model for heat loss is presented. A previously-developed model for radiative heat loss is combined with conductive heat loss theory and a new model for convective heat loss. Theoretical calculations are developed and theoretical energy consumption of the polysilicon deposition process is obtained. The model is validated by comparison with experimental results obtained using a laboratory-scale CVD reactor. Finally, the model is used to calculate heat consumption in a 36-rod industrial reactor; the energy consumption due to convective heat loss per kilogram of polysilicon produced is calculated to be 22-30 kWh/kg along a deposition process.

Azotobacter vinelandii nitrogenase: “Kinetics of nif gene expression and insights into the roles of FdxN and NifQ in FeMo-co biosynthesis”

The Molybdenum-nitrogenase is responsible for most biological nitrogen fixation activity (BNF) in the biosphere. Due to its great agronomical importance, it has been the subject of profound genetic and biochemical studies. The Mo nitrogenase carries at its active site a unique iron-molybdenum cofactor (FeMoco) that consists of an inorganic 7 Fe, 1 Mo, 1 C, 9 S core coordinated to the organic acid homocitrate. Biosynthesis of FeMo-co occurs outside nitrogenase through a complex and highly regulated pathway involving proteins acting as molecular scaffolds, metallocluster carriers or enzymes that provide substrates in appropriate chemical forms. Specific expression regulatory factors tightly control the accumulation levels of all these other components. Insertion of FeMo-co into a P-cluster containing apo-NifDK polypeptide results in nitrogenase reconstitution. Investigation of FeMo-co biosynthesis has uncovered new radical chemistry reactions and new roles for Fe-S clusters in biology.

Monte Carlo Neutronics and Thermal-hydraulics coupling applied to Fast Reactors

Un escenario habitualmente considerado para el uso sostenible y prolongado de la energía nuclear contempla un parque de reactores rápidos refrigerados por metales líquidos (LMFR) dedicados al reciclado de Pu y la transmutación de actínidos minoritarios (MA). Otra opción es combinar dichos reactores con algunos sistemas subcríticos asistidos por acelerador (ADS), exclusivamente destinados a la eliminación de MA. El diseño y licenciamiento de estos reactores innovadores requiere herramientas computacionales prácticas y precisas, que incorporen el conocimiento obtenido en la investigación experimental de nuevas configuraciones de reactores, materiales y sistemas. A pesar de que se han construido y operado un cierto número de reactores rápidos a nivel mundial, la experiencia operacional es todavía reducida y no todos los transitorios se han podido entender completamente. Por tanto, los análisis de seguridad de nuevos LMFR están basados fundamentalmente en métodos deterministas, al contrario que las aproximaciones modernas para reactores de agua ligera (LWR), que se benefician también de los métodos probabilistas. La aproximación más usada en los estudios de seguridad de LMFR es utilizar una variedad de códigos, desarrollados a base de distintas teorías, en busca de soluciones integrales para los transitorios e incluyendo incertidumbres. En este marco, los nuevos códigos para cálculos de mejor estimación ("best estimate") que no incluyen aproximaciones conservadoras, son de una importancia primordial para analizar estacionarios y transitorios en reactores rápidos. Esta tesis se centra en el desarrollo de un código acoplado para realizar análisis realistas en reactores rápidos críticos aplicando el método de Monte Carlo. Hoy en día, dado el mayor potencial de recursos computacionales, los códigos de transporte neutrónico por Monte Carlo se pueden usar de manera práctica para realizar cálculos detallados de núcleos completos, incluso de elevada heterogeneidad material. Además, los códigos de Monte Carlo se toman normalmente como referencia para los códigos deterministas de difusión en multigrupos en aplicaciones con reactores rápidos, porque usan secciones eficaces punto a punto, un modelo geométrico exacto y tienen en cuenta intrínsecamente la dependencia angular de flujo. En esta tesis se presenta una metodología de acoplamiento entre el conocido código MCNP, que calcula la generación de potencia en el reactor, y el código de termohidráulica de subcanal COBRA-IV, que obtiene las distribuciones de temperatura y densidad en el sistema. COBRA-IV es un código apropiado para aplicaciones en reactores rápidos ya que ha sido validado con resultados experimentales en haces de barras con sodio, incluyendo las correlaciones más apropiadas para metales líquidos. En una primera fase de la tesis, ambos códigos se han acoplado en estado estacionario utilizando un método iterativo con intercambio de archivos externos. El principal problema en el acoplamiento neutrónico y termohidráulico en estacionario con códigos de Monte Carlo es la manipulación de las secciones eficaces para tener en cuenta el ensanchamiento Doppler cuando la temperatura del combustible aumenta. Entre todas las opciones disponibles, en esta tesis se ha escogido la aproximación de pseudo materiales, y se ha comprobado que proporciona resultados aceptables en su aplicación con reactores rápidos. Por otro lado, los cambios geométricos originados por grandes gradientes de temperatura en el núcleo de reactores rápidos resultan importantes para la neutrónica como consecuencia del elevado recorrido libre medio del neutrón en estos sistemas. Por tanto, se ha desarrollado un módulo adicional que simula la geometría del reactor en caliente y permite estimar la reactividad debido a la expansión del núcleo en un transitorio. éste módulo calcula automáticamente la longitud del combustible, el radio de la vaina, la separación de los elementos de combustible y el radio de la placa soporte en función de la temperatura. éste efecto es muy relevante en transitorios sin inserción de bancos de parada. También relacionado con los cambios geométricos, se ha implementado una herramienta que, automatiza el movimiento de las barras de control en busca d la criticidad del reactor, o bien calcula el valor de inserción axial las barras de control. Una segunda fase en la plataforma de cálculo que se ha desarrollado es la simulació dinámica. Puesto que MCNP sólo realiza cálculos estacionarios para sistemas críticos o supercríticos, la solución más directa que se propone sin modificar el código fuente de MCNP es usar la aproximación de factorización de flujo, que resuelve por separado la forma del flujo y la amplitud. En este caso se han estudiado en profundidad dos aproximaciones: adiabática y quasiestática. El método adiabático usa un esquema de acoplamiento que alterna en el tiempo los cálculos neutrónicos y termohidráulicos. MCNP calcula el modo fundamental de la distribución de neutrones y la reactividad al final de cada paso de tiempo, y COBRA-IV calcula las propiedades térmicas en el punto intermedio de los pasos de tiempo. La evolución de la amplitud de flujo se calcula resolviendo las ecuaciones de cinética puntual. Este método calcula la reactividad estática en cada paso de tiempo que, en general, difiere de la reactividad dinámica que se obtendría con la distribución de flujo exacta y dependiente de tiempo. No obstante, para entornos no excesivamente alejados de la criticidad ambas reactividades son similares y el método conduce a resultados prácticos aceptables. Siguiendo esta línea, se ha desarrollado después un método mejorado para intentar tener en cuenta el efecto de la fuente de neutrones retardados en la evolución de la forma del flujo durante el transitorio. El esquema consiste en realizar un cálculo cuasiestacionario por cada paso de tiempo con MCNP. La simulación cuasiestacionaria se basa EN la aproximación de fuente constante de neutrones retardados, y consiste en dar un determinado peso o importancia a cada ciclo computacial del cálculo de criticidad con MCNP para la estimación del flujo final. Ambos métodos se han verificado tomando como referencia los resultados del código de difusión COBAYA3 frente a un ejercicio común y suficientemente significativo. Finalmente, con objeto de demostrar la posibilidad de uso práctico del código, se ha simulado un transitorio en el concepto de reactor crítico en fase de diseño MYRRHA/FASTEF, de 100 MW de potencia térmica y refrigerado por plomo-bismuto. ABSTRACT Long term sustainable nuclear energy scenarios envisage a fleet of Liquid Metal Fast Reactors (LMFR) for the Pu recycling and minor actinides (MAs) transmutation or combined with some accelerator driven systems (ADS) just for MAs elimination. Design and licensing of these innovative reactor concepts require accurate computational tools, implementing the knowledge obtained in experimental research for new reactor configurations, materials and associated systems. Although a number of fast reactor systems have already been built, the operational experience is still reduced, especially for lead reactors, and not all the transients are fully understood. The safety analysis approach for LMFR is therefore based only on deterministic methods, different from modern approach for Light Water Reactors (LWR) which also benefit from probabilistic methods. Usually, the approach adopted in LMFR safety assessments is to employ a variety of codes, somewhat different for the each other, to analyze transients looking for a comprehensive solution and including uncertainties. In this frame, new best estimate simulation codes are of prime importance in order to analyze fast reactors steady state and transients. This thesis is focused on the development of a coupled code system for best estimate analysis in fast critical reactor. Currently due to the increase in the computational resources, Monte Carlo methods for neutrons transport can be used for detailed full core calculations. Furthermore, Monte Carlo codes are usually taken as reference for deterministic diffusion multigroups codes in fast reactors applications because they employ point-wise cross sections in an exact geometry model and intrinsically account for directional dependence of the ux. The coupling methodology presented here uses MCNP to calculate the power deposition within the reactor. The subchannel code COBRA-IV calculates the temperature and density distribution within the reactor. COBRA-IV is suitable for fast reactors applications because it has been validated against experimental results in sodium rod bundles. The proper correlations for liquid metal applications have been added to the thermal-hydraulics program. Both codes are coupled at steady state using an iterative method and external files exchange. The main issue in the Monte Carlo/thermal-hydraulics steady state coupling is the cross section handling to take into account Doppler broadening when temperature rises. Among every available options, the pseudo materials approach has been chosen in this thesis. This approach obtains reasonable results in fast reactor applications. Furthermore, geometrical changes caused by large temperature gradients in the core, are of major importance in fast reactor due to the large neutron mean free path. An additional module has therefore been included in order to simulate the reactor geometry in hot state or to estimate the reactivity due to core expansion in a transient. The module automatically calculates the fuel length, cladding radius, fuel assembly pitch and diagrid radius with the temperature. This effect will be crucial in some unprotected transients. Also related to geometrical changes, an automatic control rod movement feature has been implemented in order to achieve a just critical reactor or to calculate control rod worth. A step forward in the coupling platform is the dynamic simulation. Since MCNP performs only steady state calculations for critical systems, the more straight forward option without modifying MCNP source code, is to use the flux factorization approach solving separately the flux shape and amplitude. In this thesis two options have been studied to tackle time dependent neutronic simulations using a Monte Carlo code: adiabatic and quasistatic methods. The adiabatic methods uses a staggered time coupling scheme for the time advance of neutronics and the thermal-hydraulics calculations. MCNP computes the fundamental mode of the neutron flux distribution and the reactivity at the end of each time step and COBRA-IV the thermal properties at half of the the time steps. To calculate the flux amplitude evolution a solver of the point kinetics equations is used. This method calculates the static reactivity in each time step that in general is different from the dynamic reactivity calculated with the exact flux distribution. Nevertheless, for close to critical situations, both reactivities are similar and the method leads to acceptable practical results. In this line, an improved method as an attempt to take into account the effect of delayed neutron source in the transient flux shape evolutions is developed. The scheme performs a quasistationary calculation per time step with MCNP. This quasistationary simulations is based con the constant delayed source approach, taking into account the importance of each criticality cycle in the final flux estimation. Both adiabatic and quasistatic methods have been verified against the diffusion code COBAYA3, using a theoretical kinetic exercise. Finally, a transient in a critical 100 MWth lead-bismuth-eutectic reactor concept is analyzed using the adiabatic method as an application example in a real system.

Evaluation of the bioaccumulation kinetics of gold nanorods in vital mammalian organs by mean of TXRF spectrometry

This work presents the first application of total-reflection X-ray fluorescence (TXRF) spectrometry, a new and powerful alternative analytical method, to evaluation of the bioaccumulation kinetics of gold nanorods (GNRs) in various tissues upon intravenous administration in mice. The analytical parameters for developed methodology by TXRF were evaluated by means of the parallel analysis of bovine liver certified reference material samples (BCR-185R) doped with 10 μg/g gold. The average values (n = 5) achieved for gold measurements in lyophilized tissue weight were as follows: recovery 99.7%, expanded uncertainty (k = 2) 7%, repeatability 1.7%, detection limit 112 ng/g, and quantification limit 370 ng/g. The GNR bioaccumulation kinetics was analyzed in several vital mammalian organs such as liver, spleen, brain, and lung at different times. Additionally, urine samples were analyzed to study the kinetics of elimination of the GNRs by this excretion route. The main achievement was clearly differentiating two kinds of behaviors. GNRs were quickly bioaccumulated by highly vascular filtration organs such as liver and spleen, while GNRs do not show a bioaccumulation rates in brain and lung for the period of time investigated. In parallel, urine also shows a lack of GNR accumulation. TXRF has proven to be a powerful, versatile, and precise analytical technique for the evaluation of GNRs content in biological systems and, in a more general way, for any kind of metallic nanoparticles.

Assessment of damage domains of the High-Temperature Engineering Test Reactor (HTTR)

This paper presents an assessment analysis of damage domains of the 30 MWth prototype High-Temperature Engineering Test Reactor (HTTR) operated by the Japan Atomic Energy Agency (JAEA). For this purpose, an in-house deterministic risk assessment computational tool was developed based on the Theory of Stimulated Dynamics (TSD). To illustrate the methodology and applicability of the developed modelling approach, assessment results of a control rod (CR) withdrawal accident during subcritical conditions are presented and compared with those obtained by the JAEA.

Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes.

Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

Estudio de la influencia de la relación C/N en la desnitrificación y metanogénesis de aguas residuales en un reactor anaeróbico de lecho fluidizado inverso

La nitrificación-desnitrificación es el proceso biológico tradicional para la remoción de nitrógeno de las aguas residuales (Ruiz G. et al., 2006a), siendo fundamental ya que contribuye a controlar la eutroficación de los cuerpos receptores. Debido al deterioro que sobre la disponibilidad de los recursos han ejercido las actividades antropogénicas, es necesario orientar el tratamiento de las aguas residuales hacia tecnologías que ofrezcan el mayor grado de sustentabilidad, planteando innovaciones en el tratamiento. El presente proyecto de tesis doctoral versa sobre el estudio de la influencia de la relación C/N en la desnitrificación y metanogénesis de aguas residuales urbanas en un reactor anaeróbico de lecho fluidizado inverso (RLFI). Previamente a la realización de las pruebas experimentales de variación de la relación C/N, se llevó a cabo la etapa de arranque del RLFI la cual se inició en modo batch, favoreciendo la formación y adhesión de biopelícula al medio de soporte utilizado (Extendosphere). Después, sobrevino la operación en modo continuo desde una carga volumétrica aplicada (CVA) de 0.5 g DQOs/L⋅d hasta alcanzar 4 g DQOs/L⋅d, carga volumétrica a la cual se logró la plena estabilización del reactor, siendo la alta variabilidad de la concentración de DQOs en el agua residual urbana de alimentación, la principal problemática que ocasionó retrasos en la estabilidad del reactor. A una CVA de 4 g DQOs/L⋅d en estado estacionario, el valor mínimo de eficiencia de remoción de DQOs fue del 32.36% y el máximo de 66.99%. En estas condiciones el porcentaje de metano presente en el biogás producido tuvo un valor medio de 85.57 ± 2.93%, siendo un valor alto comparado con otros porcentajes de metano encontrados en la digestión anaerobia de aguas residuales urbanas. El YCH4 tuvo un valor medio de 0.316 ± 0.110 LCH4/g DQOrem⋅día. Los porcentajes de metanización variaron en el rango de 20.50 a 100%, registrándose un valor medio de 73.42 ± 25.63%. La considerable variabilidad en el porcentaje de metanización se debió principalmente a que se presentaron eventos de lavado de soporte colonizado, lo cual propició que las actividades metabólicas fueran orientadas hacia formación de biopelícula (anabolismo) en vez de estar dirigidas hacia producción de metano (catabolismo). En relación a los ensayos con variación de la relación C/N, se manejaron relaciones DQOs/N-NO3 en el rango de 1.65 a 21.1 g DQOs/g N-NO3. La tasa de remoción anaerobia de DQOs se incrementó con la concentración de sustrato en una relación casi lineal, ajustándose a una cinética de primer orden, lo que regularmente se presenta a concentraciones bajas de sustrato. La eficiencia del proceso de desnitrificación fue por lo regular alta, incrementándose ligeramente con la concentración de DQOs en el influente, con valores en el rango de 73.8 a 99.1%. Por otra parte, la tasa de remoción por metanogénesis se incrementó con la concentración relativa de sustrato (es decir, a mayores relaciones DQOs/N-NO3), siendo más sensitiva la metanogénesis a la concentración relativa de sustrato que la desnitrificación. Conforme aumentó la relación DQOs/N-NO3, la desnitrificación, de ser la ruta metabólica principal de utilización de la materia orgánica (comparada con la metanización), empezó a combinarse con la metanización. De manera evidente, a las relaciones DQOs/N-NO3 probadas, se manifestaron más las actividades desnitrificantes, quedando reflejadas por el alto porcentaje de utilización de la DQOs removida hacia la desnitrificación. La relación experimental DQOs/N-NO3 a la cual se pudiera haber cumplido con el requerimiento de materia orgánica (en términos de DQOs) para la desnitrificación de nitratos en las aguas residuales urbanas tratadas resultó aproximadamente ser igual a 7.1 g DQOs/g N-NO3. A una CVA de 4 g DQOs/L⋅d, se obtuvo un diámetro promedio máximo de soporte colonizado igual a 266.106 ± 69.279 μm aunque, hay que indicarlo, se presentaron fluctuaciones, las cuales se reflejaron también en el espesor de la biopelícula, el cual tuvo un valor máximo de 50.099 μm y un valor promedio de 37.294 ± 11.199 μm. Estas fluctuaciones pudieron deberse a la existencia de corrientes preferenciales dentro del reactor, las cuales no permitieron un acceso equitativo del sustrato a todo el lecho. Nitrification-denitrification is the traditional biological process for nitrogen removal from wastewaters (Ruiz G. et al., 2006a), being fundamental since it contributes to control the eutrophication of the receiving waters. Due to the deterioration that on the availability of the aquatic resources the anthropogenic activities have exerted, it is necessary to orient the treatment of wastewaters towards technologies that offer the greater degree of sustainability, raising innovations in the treatment. This work studied the influence of C/N ratio on denitrification and methanogenesis of urban wastewaters in an inverse fluidized bed reactor (IFBR). Previously to the accomplishment of the experimental tests with variation of C/N ratio, the start up of the IFBR was carried out in batch way, encouraging the formation and adhesion of biofilm to Extendosphere, which it was used as support. The operation in continuous way carried out from an organic loading rate (OLR) of 0.5 g CODs/L ∙ d to 4 g CODs/L ∙ d, when the steady-state was reached. The high variability of the CODs of the urban wastewaters caused delays in the stability of the reactor. Once stationary state was reached, the removal efficiency of CODs ranged from 32.36 to 66.99% to 4 g CODs/L ∙ d. In these conditions the percentage of methane in produced biogas had an average value of 85.57 ± 2.93%, being a high value compared with other studies treating anaerobically urban wastewaters. The YCH4 had an average value of 0.316 ± 0.110 LCH4/g CODrem ∙ d. The percentage of methanisation ranged from 20.50 to 100%, with an average value of 73.42 ± 25.63%. The considerable variability in the methanisation percentage occurred mainly due events of wash-out of colonized support, which caused that the metabolic activities were oriented towards formation of biofilm (anabolism) instead of methane production (catabolism). Concerning the tests with variation of C/N ratio, CODs/NO3-N ratios from 1.65 to 21.1 g CODs/g NO3-N were proved. The CODs anaerobic removal rate increased with the substrate concentration in an almost linear relation, adjusting to a kinetic of first order, which regularly appears to low concentrations of substrate. Efficiency of the denitrification process was regularly high, and it increased slightly with the CODs concentration in the influent, ranging from 73.8 to 99.1%. On the other hand, the CODs removal rate by methanogenesis increased with the substrate relative concentration (e.g., to greater CODs/NO3-N ratios), being more sensitive the methanogenesis to the substrate relative concentration that the denitrification. When the CODs/NO3-N ratio increased, the denitrification, of being the main metabolic route of use of the organic matter (compared with the methanogenesis), began to be combined with the methanogenesis. Definitively, to the proven CODs/NO3-N ratios the denitrification processes were more pronounced, being reflected by the high percentage of use of the removed CODs towards denitrification. The experimental CODs/NO3-N ratio to which it was possible to have been fulfilled the requirement of organic matter (in terms of CODs) for the denitrification of nitrates in urban wastewaters turned out to be approximately 7.1 g CODs/g NO3-N. It was obtained a maximum average diameter of colonized support of 266.106 ± 69.279 μm to 4 g CODs/L ∙ d, although it is necessary to indicate that appeared fluctuations in the thickness of biofilm, which had a maximum value of 50.099 μm and an average value of 37.294 ± 11.199 μm. These fluctuations could be due to the existence of preferential currents within the reactor, which did not allow an equitable access of the substrate to all the bed.

Simulating the remote handling of the blanket segments in DEMO fusion reactor with thermo-mechanical meshfree multibody dynamics

The present study shows a first approach to the simulation of the remote handling oper- ation which takes into account the thermal and flexible behavior of the blanket segments and its implications on the remote handling equipment, in order to validate and improve its design.

Biomass Gasification in a Fluidized Bed Reactor: Effect of Temperature, Stoichiometric Ratio and Biomass Type

This paper investigates the gasification of two biomass types (pine wood and olive stones) in a laboratory scale bubbling fluidized bed reactor, in order to evaluate comparatively their potential in the production of syngas.

Deposition reactors for solar grade silicon: a comparative thermal analysis of a Siemens reactor and a fluidized bed reactor

Polysilicon production costs contribute approximately to 25-33% of the overall cost of the solar panels and a similar fraction of the total energy invested in their fabrication. Understanding the energy losses and the behaviour of process temperature is an essential requirement as one moves forward to design and build large scale polysilicon manufacturing plants. In this paper we present thermal models for two processes for poly production, viz., the Siemens process using trichlorosilane (TCS) as precursor and the fluid bed process using silane (monosilane, MS).We validate the models with some experimental measurements on prototype laboratory reactors relating the temperature profiles to product quality. A model sensitivity analysis is also performed, and the efects of some key parameters such as reactor wall emissivity, gas distributor temperature, etc., on temperature distribution and product quality are examined. The information presented in this paper is useful for further understanding of the strengths and weaknesses of both deposition technologies, and will help in optimal temperature profiling of these systems aiming at lowering production costs without compromising the solar cell quality.