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.

Core physics and safety analysis of Generation-IV Sodium Fast Reactors using existing and newly developed computational tools

El futuro de la energía nuclear de fisión dependerá, entre otros factores, de la capacidad que las nuevas tecnologías demuestren para solventar los principales retos a largo plazo que se plantean. Los principales retos se pueden resumir en los siguientes aspectos: la capacidad de proporcionar una solución final, segura y fiable a los residuos radiactivos; así como dar solución a la limitación de recursos naturales necesarios para alimentar los reactores nucleares; y por último, una mejora robusta en la seguridad de las centrales que en definitiva evite cualquier daño potencial tanto en la población como en el medio ambiente como consecuencia de cualquier escenario imaginable o más allá de lo imaginable. Siguiendo estas motivaciones, la Generación IV de reactores nucleares surge con el compromiso de proporcionar electricidad de forma sostenible, segura, económica y evitando la proliferación de material fisible. Entre los sistemas conceptuales que se consideran para la Gen IV, los reactores rápidos destacan por su capacidad potencial de transmutar actínidos a la vez que permiten una utilización óptima de los recursos naturales. Entre los refrigerantes que se plantean, el sodio parece una de las soluciones más prometedoras. Como consecuencia, esta tesis surgió dentro del marco del proyecto europeo CP-ESFR con el principal objetivo de evaluar la física de núcleo y seguridad de los reactores rápidos refrigerados por sodio, al tiempo que se desarrollaron herramientas apropiadas para dichos análisis. Efectivamente, en una primera parte de la tesis, se abarca el estudio de la física del núcleo de un reactor rápido representativo, incluyendo el análisis detallado de la capacidad de transmutar actínidos minoritarios. Como resultado de dichos análisis, se publicó un artículo en la revista Annals of Nuclear Energy [96]. Por otra parte, a través de un análisis de un hipotético escenario nuclear español, se evalúo la disponibilidad de recursos naturales necesarios en el caso particular de España para alimentar una flota específica de reactores rápidos, siguiendo varios escenarios de demanda, y teniendo en cuenta la capacidad de reproducción de plutonio que tienen estos sistemas. Como resultado de este trabajo también surgió una publicación en otra revista científica de prestigio internacional como es Energy Conversion and Management [97]. Con objeto de realizar esos y otros análisis, se desarrollaron diversos modelos del núcleo del ESFR siguiendo varias configuraciones, y para diferentes códigos. Por otro lado, con objeto de poder realizar análisis de seguridad de reactores rápidos, son necesarias herramientas multidimensionales de alta fidelidad específicas para reactores rápidos. Dichas herramientas deben integrar fenómenos relacionados con la neutrónica y con la termo-hidráulica, entre otros, mediante una aproximación multi-física. Siguiendo este objetivo, se evalúo el código de difusión neutrónica ANDES para su aplicación a reactores rápidos. ANDES es un código de resolución nodal que se encuentra implementado dentro del sistema COBAYA3 y está basado en el método ACMFD. Por lo tanto, el método ACMFD fue sometido a una revisión en profundidad para evaluar su aptitud para la aplicación a reactores rápidos. Durante ese proceso, se identificaron determinadas limitaciones que se discutirán a lo largo de este trabajo, junto con los desarrollos que se han elaborado e implementado para la resolución de dichas dificultades. Por otra parte, se desarrolló satisfactoriamente el acomplamiento del código neutrónico ANDES con un código termo-hidráulico de subcanales llamado SUBCHANFLOW, desarrollado recientemente en el KIT. Como conclusión de esta parte, todos los desarrollos implementados son evaluados y verificados. En paralelo con esos desarrollos, se calcularon para el núcleo del ESFR las secciones eficaces en multigrupos homogeneizadas a nivel nodal, así como otros parámetros neutrónicos, mediante los códigos ERANOS, primero, y SERPENT, después. Dichos parámetros se utilizaron más adelante para realizar cálculos estacionarios con ANDES. Además, como consecuencia de la contribución de la UPM al paquete de seguridad del proyecto CP-ESFR, se calcularon mediante el código SERPENT los parámetros de cinética puntual que se necesitan introducir en los típicos códigos termo-hidráulicos de planta, para estudios de seguridad. En concreto, dichos parámetros sirvieron para el análisis del impacto que tienen los actínidos minoritarios en el comportamiento de transitorios. Concluyendo, la tesis presenta una aproximación sistemática y multidisciplinar aplicada al análisis de seguridad y comportamiento neutrónico de los reactores rápidos de sodio de la Gen-IV, usando herramientas de cálculo existentes y recién desarrolladas ad' hoc para tal aplicación. Se ha empleado una cantidad importante de tiempo en identificar limitaciones de los métodos nodales analíticos en su aplicación en multigrupos a reactores rápidos, y se proponen interesantes soluciones para abordarlas. ABSTRACT The future of nuclear reactors will depend, among other aspects, on the capability to solve the long-term challenges linked to this technology. These are the capability to provide a definite, safe and reliable solution to the nuclear wastes; the limitation of natural resources, needed to fuel the reactors; and last but not least, the improved safety, which would avoid any potential damage on the public and or environment as a consequence of any imaginable and beyond imaginable circumstance. Following these motivations, the IV Generation of nuclear reactors arises, with the aim to provide sustainable, safe, economic and proliferationresistant electricity. Among the systems considered for the Gen IV, fast reactors have a representative role thanks to their potential capacity to transmute actinides together with the optimal usage of natural resources, being the sodium fast reactors the most promising concept. As a consequence, this thesis was born in the framework of the CP-ESFR project with the generic aim of evaluating the core physics and safety of sodium fast reactors, as well as the development of the approppriated tools to perform such analyses. Indeed, in a first part of this thesis work, the main core physics of the representative sodium fast reactor are assessed, including a detailed analysis of the capability to transmute minor actinides. A part of the results obtained have been published in Annals of Nuclear Energy [96]. Moreover, by means of the analysis of a hypothetical Spanish nuclear scenario, the availability of natural resources required to deploy an specific fleet of fast reactor is assessed, taking into account the breeding properties of such systems. This work also led to a publication in Energy Conversion and Management [97]. In order to perform those and other analyses, several models of the ESFR core were created for different codes. On the other hand, in order to perform safety studies of sodium fast reactors, high fidelity multidimensional analysis tools for sodium fast reactors are required. Such tools should integrate neutronic and thermal-hydraulic phenomena in a multi-physics approach. Following this motivation, the neutron diffusion code ANDES is assessed for sodium fast reactor applications. ANDES is the nodal solver implemented inside the multigroup pin-by-pin diffusion COBAYA3 code, and is based on the analytical method ACMFD. Thus, the ACMFD was verified for SFR applications and while doing so, some limitations were encountered, which are discussed through this work. In order to solve those, some new developments are proposed and implemented in ANDES. Moreover, the code was satisfactorily coupled with the thermal-hydraulic code SUBCHANFLOW, recently developed at KIT. Finally, the different implementations are verified. In addition to those developments, the node homogenized multigroup cross sections and other neutron parameters were obtained for the ESFR core using ERANOS and SERPENT codes, and employed afterwards by ANDES to perform steady state calculations. Moreover, as a result of the UPM contribution to the safety package of the CP-ESFR project, the point kinetic parameters required by the typical plant thermal-hydraulic codes were computed for the ESFR core using SERPENT, which final aim was the assessment of the impact of minor actinides in transient behaviour. All in all, the thesis provides a systematic and multi-purpose approach applied to the assessment of safety and performance parameters of Generation-IV SFR, using existing and newly developed analytical tools. An important amount of time was employed in identifying the limitations that the analytical nodal diffusion methods present when applied to fast reactors following a multigroup approach, and interesting solutions are proposed in order to overcome them.

Monte Carlo -reaktorifysiikkalaskennan ja laskennallisen virtausmekaniikan kytkentä kuulakekoreaktorissa

Monte Carlo -reaktorifysiikkakoodit nykyisin käytettävissä olevilla laskentatehoilla tarjoavat mielenkiintoisen tavan reaktorifysiikan ongelmien ratkaisuun. Neljännen sukupolven ydinreaktoreissa käytettävät uudet rakenteet ja materiaalit ovat haasteellisia nykyisiin reaktoreihin suunnitelluille laskentaohjelmille. Tässä työssä Monte Carlo -reaktorifysiikkakoodi ja CFD-koodi yhdistetään kytkettyyn laskentaan kuulakekoreaktorissa, joka on yksi korkealämpötilareaktorityyppi. Työssä käytetty lähestymistapa on uutta maailmankin mittapuussa ajateltuna.

Eksotermiset kemialliset reaktiot ydinreaktoreissa

Neljännen sukupolven reaktoreissa käytetään uusia teknisiä ratkaisuja ja uudenlaisia materiaaleja, joten myös niiden turvallisuuskriteerien laatimiseen tarvitaan uusia näkökulmia. Tällä hetkellä kehitetäänkin teknologianeutraaleja turvallisuuskriteerejä, joista voitaisiin johtaa jokaiselle uudelle reaktorikonseptille reaktorin erityispiirteet huomioivat teknologiaspesifit turvallisuuskriteerit. Näin pystytään takaamaan turvallisuuden korkea taso kaikissa uusissa reaktoreissa. Eksotermiset eli lämpöä vapauttavat kemialliset reaktiot muodostavat merkittävän uhan ydinvoimalaitosten turvallisuudelle. Tutkimalla nykyisin käytössä olevia turvallisuuskriteerejä sekä kehitteillä olevia teknologianeutraaleja turvallisuuskriteerejä voitiin havaita, että eksotermiset kemialliset reaktiot on niissä huomioitu hyvin, mutta ei kovin systemaattisesti. Tämän tutkielman tavoitteena oli pohtia, kuinka eksotermiset kemialliset reaktiot voitaisiin huomioida systemaattisemmin teknologianeutraaleissa turvallisuuskriteereissä. Johtopäätöksenä on, että epätoivottujen eksotermisten kemiallisten reaktioiden tapahtuminen tulisi ensisijaisesti pyrkiä estämään, mutta jos tällainen reaktio kuitenkin tapahtuu, tulisi sen seurauksia lieventää. Eksotermisten kemiallisten reaktioiden tapahtuminen pystytään estämään, jos eksotermisesti reagoivia aineita ei ole tai ne pystytään pitämään erillään toisistaan, tai jos lämpötilat saadaan pidettyä riittävän alhaisina. Tutkielman toisena tavoitteena oli tarkastella onnettomuusskenaarioita, jotka voisivat johtaa eksotermisiin kemiallisiin reaktioihin erityisesti neljännen sukupolven reaktoreissa. Tätä varten tutkitaan kirjallisuuden avulla joidenkin reaktorimateriaalien kemiallisia ominaisuuksia sekä muutamia neljännen sukupolven reaktoreja. Kirjallisuuden avulla tarkastellaan myös muutamaa sellaista ydinvoimalaitosonnettomuutta, joissa eksotermiset kemialliset reaktiot ovat olleet merkittävässä roolissa.

New Type Nuclear Reactors (NETNUC) 2008-2011 Final Report

This report summarizes the work done by a consortium consisting of Lappeenranta University of Technology, Aalto University and VTT Technical Research Centre of Finland in the New Type Nuclear Reactors (NETNUC) project during 2008–2011. The project was part of the Sustainable Energy (SusEn) research programme of the Academy of Finland. A wide range of generation IV nuclear technologies were studied during the project and the research consisted of multiple tasks. This report contains short articles summarizing the results of the individual tasks. In addition, the publications produced and the persons involved in the project are listed in the appendices.

Sulasuolareaktorit

Sulasuolareaktori on ydinreaktorityyppi, jota kehitettiin ensimmäisen kerran Yhdysvalloissa 1940-luvulta 1970-luvulle. Tänä aikana sulasuolatekniikkaa tutkittiin muun muassa kahden koereaktorin avulla. Vuosikymmenten hiljaiselon jälkeen kiinnostus konseptia kohtaan heräsi uudelleen 2000-luvun alussa, kun Generation IV International Forum asetti sulasuolareaktorin yhdeksi mahdolliseksi neljännen sukupolven ydinreaktorityypiksi. Sulasuolareaktori poikkeaa merkittävästi nykyisin käytössä olevista tehoreaktoreista, joiden käyttämä polttoaine on sijoitettu kiinteinä nippuina reaktorin sydämeen. Sulasuolareaktorissa polttoaine on liuotettu sulaan suolaseokseen, joka kiertää koko primääripiirissä. Suolaseos toimii siis sekä fissiilinä polttoaineena että lämmönsiirron väliaineena. Reaktorin sydämessä on hidastimena grafiittia, ja polttoainesuola saavuttaa kriittisyyden vain grafiittimoderaattorin läpi kulkiessaan. Sulasuolareaktoreihin sisältyy monia mielenkiintoisia ominaisuuksia, kuten polttoaineen käynninaikainen jälleenkäsittely sekä kevytvesireaktorien käytetyn polttoaineen kierrättäminen. Konseptin kehittäminen vaatii kuitenkin huomattavan määrän teoreettista ja kokeellista tutkimustyötä, joten sulasuolareaktoreita ei näillä näkymin odoteta olevan kaupallisessa tuotannossa vielä lähitulevaisuudessa.

Neutronics analyses for fast spectrum nuclear systems and scenario studies for advanced nuclear fuel cycles

The present PhD thesis summarizes the three-years study about the neutronic investigation of a new concept nuclear reactor aiming at the optimization and the sustainable management of nuclear fuel in a possible European scenario. A new generation nuclear reactor for the nuclear reinassance is indeed desired by the actual industrialized world, both for the solution of the energetic question arising from the continuously growing energy demand together with the corresponding reduction of oil availability, and the environment question for a sustainable energy source free from Long Lived Radioisotopes and therefore geological repositories. Among the Generation IV candidate typologies, the Lead Fast Reactor concept has been pursued, being the one top rated in sustainability. The European Lead-cooled SYstem (ELSY) has been at first investigated. The neutronic analysis of the ELSY core has been performed via deterministic analysis by means of the ERANOS code, in order to retrieve a stable configuration for the overall design of the reactor. Further analyses have been carried out by means of the Monte Carlo general purpose transport code MCNP, in order to check the former one and to define an exact model of the system. An innovative system of absorbers has been conceptualized and designed for both the reactivity compensation and regulation of the core due to cycle swing, as well as for safety in order to guarantee the cold shutdown of the system in case of accident. Aiming at the sustainability of nuclear energy, the steady-state nuclear equilibrium has been investigated and generalized into the definition of the ``extended'' equilibrium state. According to this, the Adiabatic Reactor Theory has been developed, together with a New Paradigm for Nuclear Power: in order to design a reactor that does not exchange with the environment anything valuable (thus the term ``adiabatic''), in the sense of both Plutonium and Minor Actinides, it is required indeed to revert the logical design scheme of nuclear cores, starting from the definition of the equilibrium composition of the fuel and submitting to the latter the whole core design. The New Paradigm has been applied then to the core design of an Adiabatic Lead Fast Reactor complying with the ELSY overall system layout. A complete core characterization has been done in order to asses criticality and power flattening; a preliminary evaluation of the main safety parameters has been also done to verify the viability of the system. Burn up calculations have been then performed in order to investigate the operating cycle for the Adiabatic Lead Fast Reactor; the fuel performances have been therefore extracted and inserted in a more general analysis for an European scenario. The present nuclear reactors fleet has been modeled and its evolution simulated by means of the COSI code in order to investigate the materials fluxes to be managed in the European region. Different plausible scenarios have been identified to forecast the evolution of the European nuclear energy production, including the one involving the introduction of Adiabatic Lead Fast Reactors, and compared to better analyze the advantages introduced by the adoption of new concept reactors. At last, since both ELSY and the ALFR represent new concept systems based upon innovative solutions, the neutronic design of a demonstrator reactor has been carried out: such a system is intended to prove the viability of technology to be implemented in the First-of-a-Kind industrial power plant, with the aim at attesting the general strategy to use, to the largest extent. It was chosen then to base the DEMO design upon a compromise between demonstration of developed technology and testing of emerging technology in order to significantly subserve the purpose of reducing uncertainties about construction and licensing, both validating ELSY/ALFR main features and performances, and to qualify numerical codes and tools.

A comparative study of Monte Carlo-coupled depletion codes applied to a Sodium Fast Reactor design loaded with minor actinides

inor actinides (MAs) transmutation is a main design objective of advanced nuclear systems such as generation IV Sodium Fast Reactors (SFRs). In advanced fuel cycles, MA contents in final high level waste packages are main contributors to short term heat production as well as to long-term radiotoxicity. Therefore, MA transmutation would have an impact on repository designs and would reduce the environment burden of nuclear energy. In order to predict such consequences Monte Carlo (MC) transport codes are used in reactor design tasks and they are important complements and references for routinely used deterministic computational tools. In this paper two promising Monte Carlo transport-coupled depletion codes, EVOLCODE and SERPENT, are used to examine the impact of MA burning strategies in a SFR core, 3600 MWth. The core concept proposal for MA loading in two configurations is the result of an optimization effort upon a preliminary reference design to reduce the reactivity insertion as a consequence of sodium voiding, one of the main concerns of this technology. The objective of this paper is double. Firstly, efficiencies of the two core configurations for MA transmutation are addressed and evaluated in terms of actinides mass changes and reactivity coefficients. Results are compared with those without MA loading. Secondly, a comparison of the two codes is provided. The discrepancies in the results are quantified and discussed.

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.

Effect of aspect ratio on thermal convection in open and closed systems

Internally heated fluids are found across the nuclear fuel cycle. In certain situations the motion of the fluid is driven by the decay heat (i.e. corium melt pools in severe accidents, the shutdown of liquid metal reactors, molten salt and the passive control of light water reactors) as well as normal operation (i.e. intermediate waste storage and generation IV reactor designs). This can in the long-term affect reactor vessel integrity or lead to localized hot spots and accumulation of solid wastes that may prompt local increases in activity. Two approaches to the modeling of internally heated convection are presented here. These are based on numerical analysis using codes developed in-house and simulations using widely available computational fluid dynamics solvers. Open and closed fluid layers at around the transition between conduction and convection of various aspect ratios are considered. We determine optimum domain aspect ratio (1:7:7 up to 1:24:24 for open systems and 5:5:1, 1:10:10 and 1:20:20 for closed systems), mesh resolutions and turbulence models required to accurately and efficiently capture the convection structures that evolve when perturbing the conductive state of the fluid layer. Note that the open and closed fluid layers we study here are bounded by a conducting surface over an insulating surface. Conclusions will be drawn on the influence of the periodic boundary conditions on the flow patterns observed. We have also examined the stability of the nonlinear solutions that we found with the aim of identifying the bifurcation sequence of these solutions en route to turbulence.

THE MICROSTRUCTURES OF ZR-EXCEL ALLOY: EFFECTS OF HEAVY ION IRRADIATION, PLASTIC DEFORMATION, THERMAL TREATMENT AND IN-SITU HEATING

Zr-Excel alloy (Zr-3.5Sn-0.8Nb-0.8Mo) is a dual phase (α + β) alloy in the as-received pressure tube condition. It has been proposed to be the pressure tube candidate material for the Generation-IV CANDU-Supercritical Water Reactor (CANDU-SCWR). In this dissertation, the effects of heavy ion irradiation, deformation and heat treatment on the microstructures of the alloy were investigated to enable us to have a better understanding of the potential in-reactor performance of this alloy. In-situ heavy ion (1 MeV) irradiation was performed to study the nucleation and evolution of dislocation loops in both α- and β-Zr. Small and dense type dislocation loops form under irradiation between 80 and 450 °C. The number density tends to saturate at ~ 0.1 dpa. Compared with the α-Zr, the defect yield is much lower in β-Zr. The stabilities of the metastable phases (β-Zr and ω-Zr) and the thermal-dynamically equilibrium phase, fcc Zr(Mo, Nb)2, under irradiation were also studied at different temperatures. Chemi-STEM elemental mapping was carried out to study the elemental redistribution caused by irradiation. The stability of these phases and the elemental redistribution are strongly dependent on irradiation temperature. In-situ time-of-flight neutron diffraction tensile and compressive tests were carried out at different temperatures to monitor lattice strain evolutions of individual grain families during these tests. The β-Zr is the strengthening phase in this alloy in the as-received plate material. Load is transferred to the β-Zr after yielding of the α-Zr grains. The temperature dependence of static strain aging and the yielding sequence of the individual grain families were discussed. Strong tensile/compressive asymmetry was observed in the {0002} grain family at room temperature. The microstructures of the sample deformed at 400 °C and the samples only subjected to heat treatment at the same temperature were characterized with TEM. Concentration of β phase stabilizers in the β grain and the morphology of β grain have significant effect on the stability of β- and ω-Zr under thermal treatment. Applied stress/strain enhances the decomposition of isothermal ω phase but suppresses α precipitation inside the β grains at high temperature. An α → ω/ZrO phase transformation was observed in the thin foils of Zr-Excel alloy and pure Zr during in-situ heating at 700 °C in TEM.

Triggering the generation of an iron(IV)-oxo compound and its reactivity toward sulfides by RuII photocatalysis

The preparation of [FeIV(O)(MePy2tacn)]2+ (2, MePy2tacn = N-methyl-N,N-bis(2-picolyl)-1,4,7-triazacyclononane) by reaction of [FeII(MePy2tacn)(solvent)]2+ (1) and PhIO in CH3CN and its full characterization are described. This compound can also be prepared photochemically from its iron(II) precursor by irradiation at 447 nm in the presence of catalytic amounts of [Ru II(bpy)3]2+ as photosensitizer and a sacrificial electron acceptor (Na2S2O8). Remarkably, the rate of the reaction of the photochemically prepared compound 2 toward sulfides increases 150-fold under irradiation, and 2 is partially regenerated after the sulfide has been consumed; hence, the process can be repeated several times. The origin of this rate enhancement has been established by studying the reaction of chemically generated compound 2 with sulfides under different conditions, which demonstrated that both light and [Ru II(bpy)3]2+ are necessary for the observed increase in the reaction rate. A combination of nanosecond time-resolved absorption spectroscopy with laser pulse excitation and other mechanistic studies has led to the conclusion that an electron transfer mechanism is the most plausible explanation for the observed rate enhancement. According to this mechanism, the in-situ-generated [RuIII(bpy)3] 3+ oxidizes the sulfide to form the corresponding radical cation, which is eventually oxidized by 2 to the corresponding sulfoxide

Method development and optimization for the determination of selenium in bean and soil samples using hydride generation electrothermal atomic absorption spectrometry

The present investigation is the first part of an initiative to prepare a regional map of the natural abundance of selenium in various areas of Brazil, based on the analysis of bean and soil samples. Continuous-flow hydride generation electrothermal atomic absorption spectrometry (HG-ET AAS) with in situ trapping on an iridium-coated graphite tube has been chosen because of the high sensitivity and relative simplicity. The microwave-assisted acid digestion for bean and soil samples was tested for complete recovery of inorganic and organic selenium compounds (selenomethionine). The reduction of Se(VI) to Se(IV) was optimized in order to guarantee that there is no back-oxidation, which is of importance when digested samples are not analyzed immediately after the reduction step. The limits of detection and quantification of the method were 30 ng L(-1) Se and 101 ng L(-1) Se, respectively, corresponding to about 3 ng g(-1) and 10 ng g(-1), respectively, in the solid samples, considering a typical dilution factor of 100 for the digestion process. The results obtained for two certified food reference materials (CRM), soybean and rice, and for a soil and sediment CRM confirmed the validity of the investigated method. The selenium content found in a number of selected bean samples varied between 5.5 +/- 0.4 ng g(-1) and 1726 +/- 55 ng g(-1), and that in soil samples varied between 113 +/- 6.5 ng g(-1) and 1692 +/- 21 ng g(-1). (C) 2011 Elsevier B.V. All rights reserved.

Small Bowel Bacterial Overgrowth is found in 71% of IBS patients and associated symptoms respond wel to Rifaximin in a phase IV trial

Background: The prevalence of small intestinal bowel bacterial overgrowth (SIBO) in patients with irritable bowel syndrome (IBS) ranges from 43% to 78% as determined by the lactulose hydrogen breath (LHBT) test. Although rifaximine, a non-absorbable antibiotic, has been able to decrease global IBS symptoms as well as bloating in placebo-controlled randomized trials, these results were not repeated in phase IV studies in daily clinical practice. Aim: To assess the prevalence of SIBO in an IBS cohort and to evaluate the treatment response in the IBS cohort affected by SIBO. Methods: Enrolled patients were diagnosed with IBS using the following criteria: fulfillment of the Rome III criteria, absence of alarm symptoms (anemia, weight loss, nocturnal symptoms etc), normal fecal calproectin, normal endoscopic workup including histology. Celiac disease was excluded by serology and/or duodenal biopsy. All patients underwent lactulose hydrogen breath testing (LHBT) for SIBO diagnosis. Patients with SIBO were treated with rifaximine tablets (400mg twice daily for 14 days). Both before and at week 6 after rifaximin treatment, patients completed a questionnaire, where the following criteria were assessed individually using 11-point Likert scales: the bloating, flatulence, abdominal pain, diarrhea, and overall well-being. Results: Hundred-fifty IBS patients were enrolled (76% female, mean age 44 ± 16 years), of whom 106 (71%) were diagnosed with SIBO and consequently treated with rifaximine. Rifaximine treatment significantly reduced the following symptoms as assessed by the symptom questionnaire: bloating (5.5 ± 2.6 before vs. 3.6 ± 2.7 after treatment, p <0.001), flatulence (5 ± 2.7 vs. 4 ± 2.7, p = 0.015), diarrhea (2.9 ± 2.4 vs. 2 ± 2.4, p = 0.005), abdominal pain (4.8 ± 2.7 vs. 3.3 ± 2.5, p <0.001) and resulted in improved overall well-being (3.9 ± 2.4 vs. 2.7 ± 2.3, p <0.001). Thirteen of the 106 treated patients were lost to follow-up (12%). The LHBT was repeated 2-4 weeks after rifaximine treatment in 65/93 (70%) patients. Eradication of SIBO was documented in 85% of all patients (55/65), whereas 15% of patients (10/65) tested positive for SIBO as determined by the LHBT testing. Conclusions: The results of our phase IV trial indicate that a high proportion of IBS patients tested positive for SIBO. IBS symptoms (bloating, flatulence, diarrhea, pain, overall well-being) were significantly diminished following a 2-week treatment with rifaximine. These results support the previous findings of randomized controlled trials that the presence of SIBO is associated with symptom generation in IBS patients and that reduction and/or elimination of SIBO may help to alleviate IBSassociated symptoms.