Performance and composition of bacterial communiies in anaerobic fluidized bed reactors for hydrogen production: Effects of organic loading rate and alkalinity

Conselho Nacional de Desenvolvimento Científico e Tecnológico (CNPq)

Performance and composition of bacterial communities in anaerobic fluidized bed reactors for hydrogen production: Effects of organic loading rate and alkalinity

This study evaluated the effects of the organic loading rate (OLR) and pH buffer addition on hydrogen production in two anaerobic fluidized bed reactors (AFBRs) operated simultaneously. The AFBRs were fed with glucose, and expanded clay was used as support material. The reactors were operated at a temperature of 30 degrees C, without the addition of a buffer (AFBR1) and with the addition of a pH buffer (AFBR2, sodium bicarbonate) for OLRs ranging from 19.0 to 140.6 kg COD m(-3) d(-1) (COD: chemical oxygen demand). The maximum hydrogen yields for AFBR1 and AFBR2 were 2.45 and 1.90 mol H-2 mol(-1) glucose (OLR of 84.3 kg COD m(-3) d(-1)), respectively. The highest hydrogen production rates were 0.95 and 0.76 L h(-1) L-1 for AFBR1 and AFBR2 (OLR of 140.6 kg COD m(-3) d(-1)), respectively. The operating conditions in AFBR1 favored the presence of such bacteria as Clostridium, while the bacteria in AFBR2 included Clostridium, Enterobacter, Klebsiella, Veillonellaceae, Chryseobacterium, Sporolactobacillus, and Burkholderiaceae. Copyright (C) 2012, Hydrogen Energy Publications, LLC. Published by Elsevier Ltd. All rights reserved.

Application of gas-cooled Accelerator Driven System (ADS) transmutation devices to sustainable nuclear energy development

The conceptual design of a pebble bed gas-cooled transmutation device is shown with the aim to evaluate its potential for its deployment in the context of the sustainable nuclear energy development, which considers high temperature reactors for their operation in cogeneration mode, producing electricity, heat and Hydrogen. As differential characteristics our device operates in subcritical mode, driven by a neutron source activated by an accelerator that adds clear safety advantages and fuel flexibility opening the possibility to reduce the nuclear stockpile producing energy from actual LWR irradiated fuel with an efficiency of 45?46%, either in the form of Hydrogen, electricity, or both.

Analysis of a Spanish energy scenario with Generation IV nuclear reactors

The advantages of fast-spectrum reactors consist not only of an efficient use of fuel through the breeding of fissile material and the use of natural or depleted uranium, but also of the potential reduction of the amount of actinides such as americium and neptunium contained in the irradiated fuel. The first aspect means a guaranteed future nuclear fuel supply. The second fact is key for high-level radioactive waste management, because these elements are the main responsible for the radioactivity of the irradiated fuel in the long term. The present study aims to analyze the hypothetical deployment of a Gen-IV Sodium Fast Reactor (SFR) fleet in Spain. A nuclear fleet of fast reactors would enable a fuel cycle strategy different than the open cycle, currently adopted by most of the countries with nuclear power. A transition from the current Gen-II to Gen-IV fleet is envisaged through an intermediate deployment of Gen-III reactors. Fuel reprocessing from the Gen-II and Gen-III Light Water Reactors (LWR) has been considered. In the so-called advanced fuel cycle, the reprocessed fuel used to produce energy will breed new fissile fuel and transmute minor actinides at the same time. A reference case scenario has been postulated and further sensitivity studies have been performed to analyze the impact of the different parameters on the required reactor fleet. The potential capability of Spain to supply the required fleet for the reference scenario using national resources has been verified. Finally, some consequences on irradiated final fuel inventory are assessed. Calculations are performed with the Monte Carlo transport-coupled depletion code SERPENT together with post-processing tools.

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.

Mixing performance evaluation of additive manufactured milli-scale reactors

The mixing performance of three passive milli-scale reactors with different geometries was investigated at different Reynolds numbers. The effects of design and operating characteristics such as mixing channel shape and volume flow rate were investigated. The main objective of this work was to demonstrate a process design method that uses on Computational Fluid Dynamics (CFD) for modeling and Additive Manufacturing (AM) technology for manufacture. The reactors were designed and simulated using SolidWorks and Fluent 15.0 software, respectively. Manufacturing of the devices was performed with an EOS M-series AM system. Step response experiments with distilled Millipore water and sodium hydroxide solution provided time-dependent concentration profiles. Villermaux-Dushman reaction experiments were also conducted for additional verification of CFD results and for mixing efficiency evaluation of the different geometries. Time-dependent concentration data and reaction evaluation showed that the performance of the AM-manufactured reactors matched the CFD results reasonably well. The proposed design method allows the implementation of new and innovative solutions, especially in the process design phase, for industrial scale reactor technologies. In addition, rapid implementation is another advantage due to the virtual flow design and due to the fast manufacturing which uses the same geometric file formats.

Effect of pH and temperature on the carbonate promoted dissolution of sodium meta-autunite

Release of uranium from Na-autunite, an artificial mineral created as a result of polyphosphate injection in the subsurface at the DOE Hanford Site, takes place during slow dissolution of the mineral structure. Stability information of the uranyl-phosphate phases is limited to conditions involving pH, temperature, and a few aqueous organic materials. The carbonate ion, which creates very strong complexes with uranium, is the predominant ion in the groundwater composition. The polyphosphate technology with the formation of autunite was identified as the most feasible remediation strategy to sequester uranium in contaminated groundwater and soil in situ. The objectives of the experimental work were (i) to quantify the effect of bicarbonate on the stability of synthetic sodium meta-autunite created as a result of uranium stabilization through polyphosphate injection, (ii) calculate the kinetic rate law parameters of the uranium release from Na-autunite during dissolution, and (iii) to compare the process parameters with those obtained for natural calcium meta-autunite. Experiments were conducted using SPTF apparatus, which consists of syringe pumps for controlling flow rate, Teflon reactors and a heating/cooling system. 0.25 grams of synthetic Na-autunite was placed in the reactor and buffer solutions with varying bicarbonate concentrations (0.0005 to 0.003 M) at different pH (6 - 11) were pumped through the reactors. Experiments were conducted at four different temperatures in the range of 5 - 60oC. It was concluded that the rate of release of uranium from synthetic Na-autunite is directly correlated to the bicarbonate concentration. The rate of release of uranium increased from 1.90 x 10-12 at pH 6 to 2.64 x 10-10 (mol m-2 s-1) at pH 11 at 23oC over the bicarbonate concentration range tested. The activation energy values were invariant with the change in the bicarbonate concentration; however, pH is shown to influence the activation energy values. Uranyl hydroxides and uranyl carbonates complexes helped accelerate the dissolution of autunite mineral.

Sodium 2-nitrocinnamate dihydrate: a one-dimensional hydrogen-bonded coordination polymer

The title compound catena-poly[aqua-mu3-2-nitrocinnamato], [Na(C9H6NO4)(H2O)2]n, the sodium salt of trans-2-nitrocinnamic acid, is a one-dimensional coordination polymer based on six-coordinate octahedral NaO6 centres comprising three facially-related monodentate carboxylate O-atom donors from separate ligands (all bridging)[Na-O, 2.4370(13)-2.5046(13)A] and three water molecules (two bridging, one monodentate) [Na-O, 2.3782(13)-2.4404(17)A]. The structure is also stabilized by intra-chain water-O-H...O(carboxylate) and O-H...O(nitro) hydrogen bonds.

Fluorescence spectrometric study on the interactions of isoprocarb and sodium 2-isopropylphenate with bovine serum albumin

The binding interaction of the pesticide Isoprocarb and its degradation product, sodium 2-isopropylphenate, with bovine serum albumin (BSA) was studied by spectrofluorimetry under simulated physiological conditions. Both Isoprocarb and sodium 2-isopropylphenate quenched the intrinsic fluorescence of BSA. This quenching proceeded via a static mechanism. The thermodynamic parameters (ΔH°, ΔS° and ΔG°) obtained from the fluorescence data measured at two different temperatures showed that the binding of Isoprocarb to BSA involved hydrogen bonds and that of sodium 2-isopropylphenate to BSA involved hydrophobic and electrostatic interactions. Synchronous fluorescence spectroscopy of the interaction of BSA with either Isoprocarb or sodium 2-isopropylphenate showed that the molecular structure of the BSA was changed significantly, which is consistent with the known toxicity of the pesticide, i.e., the protein is denatured. The sodium 2-isopropylphenate, was estimated to be about 4–5 times more toxic than its parent, Isoprocarb. Synchronous fluorescence spectroscopy and the resolution of the three-way excitation–emission fluorescence spectra by the PARAFAC method extracted the relative concentration profiles of BSA, Isoprocab and sodium 2-isopropylphenate as a function of the added sodium 2-isopropylphenate. These profiles showed that the degradation product, sodium 2-isopropylphenate, displaced the pesticide in a competitive reaction with the BSA protein.

Thermoanalytical studies of natural potassium, sodium and ammonium alunites

Dynamic and controlled rate thermal analysis (CRTA) has been used to characterise alunites of formula [M(Al)3(SO4)2(OH)6 ] where M+ is the cations K+, Na+ or NH4+. Thermal decomposition occurs in a series of steps. (a) dehydration, (b) well defined dehydroxylation and (c) desulphation. CRTA offers a better resolution and a more detailed interpretation of water formation processes via approaching equilibrium conditions of decomposition through the elimination of the slow transfer of heat to the sample as a controlling parameter on the process of decomposition. Constant-rate decomposition processes of water formation reveal the subtle nature of dehydration and dehydroxylation.

Low sodium haemodialysis reduces interdialytic fluid consumption but paradoxically increases post-dialysis thirst

Background Interdialytic weight gain (IDWG) can be reduced by lowering the dialysate sodium concentration ([Na]) in haemodialysis patients. It has been assumed that this is because thirst is reduced, although this has been difficult to prove. We compared thirst patterns in stable haemodialysis patients with high and low IDWG using a novel technique and compared the effect of low sodium dialysis (LSD) with normal sodium dialysis (NSD). Methods Eight patients with initial high IDWG and seven with low IDWG completed hourly visual analogue ratings of thirst using a modified palmtop computer during the dialysis day and the interdialytic day. The dialysate [Na] was progressively reduced by up to 5 mmol/l over five treatments. Dialysis continued at the lowest attained [Na] for 2 weeks and the measurements were repeated. The dialysate [Na] then returned to baseline and the process was repeated. Results Baseline interdialytic day mean thirst was higher than the dialysis day mean for the high IDWG group (49.9±14.0 vs 36.2±16.6) and higher than the low weight gain group (49.9±14.0 vs 34.1±14.6). This trend persisted on LSD, but there was a pronounced increase in post-dialysis thirst scores for both groups (high IDWG: 46±13 vs 30±21; low IDWG: 48±24 vs 33±18). The high IDWG group demonstrated lower IDWG during LSD than NSD (2.23±0.98 vs 2.86±0.38 kg; P<0.05). Conclusions Our results indicate that patients with high IDWG experience more intense feelings of thirst on the interdialytic day. LSD reduces their IDWG, but paradoxically increases thirst in the immediate post-dialysis period.