Isotope effect in proton-transfer and association-dissociation reactions for butylamine by ultrasonic relaxation methods

Ultrasonic absorption coefficients were measured for butylamine in heavy water (D2O) in the frequency range from 0.8 to 220 MHz and at concentrations from 0.0278 to 2.5170 mol dm(-3) at 25 degrees C; two kinds of relaxation processes were observed. One was found in relatively dilute solutions (up to 0.5 mol dm(-3)), which was attributed to the hydrolysis of butylamine. In order to compare the results, absorption measurements were also carried out in light water (H2O). The rate and thermodynamic parameters were determined from the concentration dependence of the relaxation frequency and the maximum absorption per wavelength. The isotope effects on the diffusion-controlled reaction were estimated and the stability of the intermediate of the hydrolysis was considered while comparing it with the results for propylamine in H2O and D2O. Another relaxation process was observed at concentrations greater than 1 mol dm(-3) in D2O. In order to examine the solution characteristics, proton NMR measurements for butylamine were also carried out in D2O. The chemical shifts for the gamma- and delta-proton in butylamine molecule indicate the existence of an aggregate. From profiles of the concentration dependence of the relaxation frequency and the maximum absorption per wavelength of sound absorption, the source of the relaxation was attributed to an association-dissociation reaction, perhaps, associated with a hydrophobic interaction. The aggregation number, the forward and reverse rate constants and the standard volume change of the reaction were determined. It was concluded from a comparison with the results in H2O that the hydrophobic interaction of butylamine in D2O is stronger than that in H2O. Also, the isotope effect on this reaction was interpreted in terms of the solvent structure.

Ultrasonic relaxation kinetics on fast deuteron transfer reaction

Propylamine has been selected to investigate the isotope effect of a fast deuteron transfer reaction by ultrasonic relaxation method. Ultrasonic absorption coefficients of propylamine in heavy water (D2O) at 25 degrees C in the concentration range from 0.0107 to 0.6300 mol dm(-3) have been measured by pulse and resonance methods over the frequency range from 0.8 to 220 MHz. A Debye-type single relaxation absorption has been observed in the solution. From the dependence of the ultrasonic relaxation parameters on the concentration and solution pH, the source of the observed relaxation has been attributed to a perturbation of the chemical equilibrium associated with the deuteron transfer reaction. The rate and equilibrium constants have been determined by the measurement of the deuteroxyl ion concentration dependence of the relaxation frequency. Also the standard volume change of the reaction has been determined from the concentration dependence of the maximum absorption per wavelength and the adiabatic compressibility has been calculated from the density and the sound velocity in the solution. These results have then been compared with those obtained for propylamine in light water (H2O). The forward rate constant is greater and the reverse rate constant is smaller in DO than in H2O. The standard volume change for deuteron transfer is greater than that for proton transfer reaction, and the adiabatic compressibility shows a similar trend. These data support an argument that there exists a stronger hydrogen bond in D2O than in H2O. The difference of the stability in the intermediate states, R-ND3+... OD- and R-NH3+... OH-, has also been considered from the results of the isotope effects.

A study of the kinetic solvent isotope effect on the destruction of microcystin-LR and geosmin using TiO2 photocatalysis

We have previously reported the effectiveness of TiO₂ photocatalysis in the destruction of species generated by cyanobacteria, specifically geosmin and microcystin-LR. In this paper we report an investigation of factors which influence the rate of the toxin destruction at the catalyst surface. A primary kinetic solvent isotope effect of approximately 1.5 was observed when the destruction was performed in a heavy water solvent. This is in contrast to previous reports of a solvent isotope effect of approximately 3, however, these studies were undertaken with a different photocatalyst material. The solvent isotope effect therefore appears to be dependent on the photocatalyst material used. The results of the study support the theory that the photocatalytic decomposition occurs on the catalyst surface rather than in the bulk of the solution. Furthermore it appears that the rate determining step is not oxygen reduction as previously reported. 

Processes influencing the destruction of microcystin-LR by TiO2 photocatalysis

We have previously reported the effectiveness of TiO₂ photocatalysis in the destruction of the cyanotoxin microcystin-LR [P.K.J. Robertson, L.A. Lawton, B. Münch, J. Rouzade, J. Chem. Soc., Chem. Commun., 4 (1997) 393; P.K.J. Robertson, L.A. Lawton, B. Münch, B.J.P.A. Cornish, J. Adv. Oxid. Technol., in press]. In this paper we report an investigation of factors which influence the rate of the toxin destruction at the catalyst surface. A primary kinetic isotope effect of approximately 3 was observed when the destruction was performed in a heavy water solvent. Hydroxylated compounds were observed as products of the destruction process. No destruction was observed when the process was investigated under a nitrogen atmosphere.

Selective deuterium ion acceleration using the Vulcan petawatt laser

We report on the successful demonstration of selective acceleration of deuterium ions by target-normal sheath acceleration (TNSA) with a high-energy petawatt laser. TNSA typically produces a multi-species ion beam that originates from the intrinsic hydrocarbon and water vapor contaminants on the target surface. Using the method first developed by Morrison et al. [Phys. Plasmas 19, 030707 (2012)], an ion beam with >99% deuterium ions and peak energy 14 MeV/nucleon is produced with a 200 J, 700 fs, > 10 20 W/cm 2 laser pulse by cryogenically freezing heavy water (D<inf>2</inf>O) vapor onto the rear surface of the target prior to the shot. Within the range of our detectors (0°-8.5°), we find laser-to-deuterium-ion energy conversion efficiency of 4.3% above 0.7 MeV/nucleon while a conservative estimate of the total beam gives a conversion efficiency of 9.4%.

Calculation of the eigenfunctions of the two-group neutron diffusion equation and application to modal decomposition of BWR instabilities

In this thesis, numerical methods aiming at determining the eigenfunctions, their adjoint and the corresponding eigenvalues of the two-group neutron diffusion equations representing any heterogeneous system are investigated. First, the classical power iteration method is modified so that the calculation of modes higher than the fundamental mode is possible. Thereafter, the Explicitly-Restarted Arnoldi method, belonging to the class of Krylov subspace methods, is touched upon. Although the modified power iteration method is a computationally-expensive algorithm, its main advantage is its robustness, i.e. the method always converges to the desired eigenfunctions without any need from the user to set up any parameter in the algorithm. On the other hand, the Arnoldi method, which requires some parameters to be defined by the user, is a very efficient method for calculating eigenfunctions of large sparse system of equations with a minimum computational effort. These methods are thereafter used for off-line analysis of the stability of Boiling Water Reactors. Since several oscillation modes are usually excited (global and regional oscillations) when unstable conditions are encountered, the characterization of the stability of the reactor using for instance the Decay Ratio as a stability indicator might be difficult if the contribution from each of the modes are not separated from each other. Such a modal decomposition is applied to a stability test performed at the Swedish Ringhals-1 unit in September 2002, after the use of the Arnoldi method for pre-calculating the different eigenmodes of the neutron flux throughout the reactor. The modal decomposition clearly demonstrates the excitation of both the global and regional oscillations. Furthermore, such oscillations are found to be intermittent with a time-varying phase shift between the first and second azimuthal modes.

Stable oxygen isotope ratios of pore waters from ODP Leg 166 sites

This study investigates the d18O of pore waters from Sites 1003 through 1007, drilled along the western margin of the Great Bahama Bank during Leg 166 of the Ocean Drilling Program. These pore waters generally show a positive correlation between d18O and the concentration of chloride. The exception to this trend is Site 1006, where the pore waters exhibit nonlinear behavior with respect to chloride. The correlation between the concentration of Cl- and d18O at most of the sites appears to be a coincidence because although the increase in Cl- is a result of diffusion from an underlying source, the increases in d18O result from the recrystallization of metastable carbonates in the presence of a geothermal gradient. The difference in behavior in the d18O of the pore water at Site 1006 is probably a result of the relative reduced rate of carbonate recrystallization at this site. The d18O of the pore waters in the upper portion of the cores shows a pattern similar to the concentration of chloride in that there is an interval of 30-50 m in which neither the d18O nor the concentration of Cl- changes. This interval is consistent with either an interval of very rapid deposition of sediment or the advection of fluid through the platform. Both the d18O and the concentration of Cl- increase toward the platform, suggesting an input of saline and isotopically heavy water from the platform surface.

Análisis Integrado de Seguridad de un accidente de SGTR en un reactor nuclear tipo PWR

El accidente de rotura de tubos de un generador de vapor (Steam Generator Tube Rupture, SGTR) en los reactores de agua a presión es uno de los transitorios más exigentes desde el punto de vista de operación. Los transitorios de SGTR son especiales, ya que podría dar lugar a emisiones radiológicas al exterior sin necesidad de daño en el núcleo previo o sin que falle la contención, ya que los SG pueden constituir una vía directa desde el reactor al medio ambiente en este transitorio. En los análisis de seguridad, el SGTR se analiza desde un punto determinista y probabilista, con distintos enfoques con respecto a las acciones del operador y las consecuencias analizadas. Cuando comenzaron los Análisis Deterministas de Seguridad (DSA), la forma de analizar el SGTR fue sin dar crédito a la acción del operador durante los primeros 30 min del transitorio, lo que suponía que el grupo de operación era capaz de detener la fuga por el tubo roto dentro de ese tiempo. Sin embargo, los diferentes casos reales de accidentes de SGTR sucedidos en los EE.UU. y alrededor del mundo demostraron que los operadores pueden emplear más de 30 minutos para detener la fuga en la vida real. Algunas metodologías fueron desarrolladas en los EEUU y en Europa para abordar esa cuestión. En el Análisis Probabilista de Seguridad (PSA), las acciones del operador se tienen en cuenta para diseñar los cabeceros en el árbol de sucesos. Los tiempos disponibles se utilizan para establecer los criterios de éxito para dichos cabeceros. Sin embargo, en una secuencia dinámica como el SGTR, las acciones de un operador son muy dependientes del tiempo disponible por las acciones humanas anteriores. Además, algunas de las secuencias de SGTR puede conducir a la liberación de actividad radiológica al exterior sin daño previo en el núcleo y que no se tienen en cuenta en el APS, ya que desde el punto de vista de la integridad de núcleo son de éxito. Para ello, para analizar todos estos factores, la forma adecuada de analizar este tipo de secuencias pueden ser a través de una metodología que contemple Árboles de Sucesos Dinámicos (Dynamic Event Trees, DET). En esta Tesis Doctoral se compara el impacto en la evolución temporal y la dosis al exterior de la hipótesis más relevantes encontradas en los Análisis Deterministas a nivel mundial. La comparación se realiza con un modelo PWR Westinghouse de tres lazos (CN Almaraz) con el código termohidráulico TRACE, con hipótesis de estimación óptima, pero con hipótesis deterministas como criterio de fallo único o pérdida de energía eléctrica exterior. Las dosis al exterior se calculan con RADTRAD, ya que es uno de los códigos utilizados normalmente para los cálculos de dosis del SGTR. El comportamiento del reactor y las dosis al exterior son muy diversas, según las diferentes hipótesis en cada metodología. Por otra parte, los resultados están bastante lejos de los límites de regulación, pese a los conservadurismos introducidos. En el siguiente paso de la Tesis Doctoral, se ha realizado un análisis de seguridad integrado del SGTR según la metodología ISA, desarrollada por el Consejo de Seguridad Nuclear español (CSN). Para ello, se ha realizado un análisis termo-hidráulico con un modelo de PWR Westinghouse de 3 lazos con el código MAAP. La metodología ISA permite la obtención del árbol de eventos dinámico del SGTR, teniendo en cuenta las incertidumbres en los tiempos de actuación del operador. Las simulaciones se realizaron con SCAIS (sistema de simulación de códigos para la evaluación de la seguridad integrada), que incluye un acoplamiento dinámico con MAAP. Las dosis al exterior se calcularon también con RADTRAD. En los resultados, se han tenido en cuenta, por primera vez en la literatura, las consecuencias de las secuencias en términos no sólo de daños en el núcleo sino de dosis al exterior. Esta tesis doctoral demuestra la necesidad de analizar todas las consecuencias que contribuyen al riesgo en un accidente como el SGTR. Para ello se ha hecho uso de una metodología integrada como ISA-CSN. Con este enfoque, la visión del DSA del SGTR (consecuencias radiológicas) se une con la visión del PSA del SGTR (consecuencias de daño al núcleo) para evaluar el riesgo total del accidente. Abstract Steam Generator Tube Rupture accidents in Pressurized Water Reactors are known to be one of the most demanding transients for the operating crew. SGTR are special transient as they could lead to radiological releases without core damage or containment failure, as they can constitute a direct path to the environment. The SGTR is analyzed from a Deterministic and Probabilistic point of view in the Safety Analysis, although the assumptions of the different approaches regarding the operator actions are quite different. In the beginning of Deterministic Safety Analysis, the way of analyzing the SGTR was not crediting the operator action for the first 30 min of the transient, assuming that the operating crew was able to stop the primary to secondary leakage within that time. However, the different real SGTR accident cases happened in the USA and over the world demonstrated that operators can took more than 30 min to stop the leakage in actual sequences. Some methodologies were raised in the USA and in Europe to cover that issue. In the Probabilistic Safety Analysis, the operator actions are taken into account to set the headers in the event tree. The available times are used to establish the success criteria for the headers. However, in such a dynamic sequence as SGTR, the operator actions are very dependent on the time available left by the other human actions. Moreover, some of the SGTR sequences can lead to offsite doses without previous core damage and they are not taken into account in PSA as from the point of view of core integrity are successful. Therefore, to analyze all this factors, the appropriate way of analyzing that kind of sequences could be through a Dynamic Event Tree methodology. This Thesis compares the impact on transient evolution and the offsite dose of the most relevant hypothesis of the different SGTR analysis included in the Deterministic Safety Analysis. The comparison is done with a PWR Westinghouse three loop model in TRACE code (Almaraz NPP), with best estimate assumptions but including deterministic hypothesis such as single failure criteria or loss of offsite power. The offsite doses are calculated with RADTRAD code, as it is one of the codes normally used for SGTR offsite dose calculations. The behaviour of the reactor and the offsite doses are quite diverse depending on the different assumptions made in each methodology. On the other hand, although the high conservatism, such as the single failure criteria, the results are quite far from the regulatory limits. In the next stage of the Thesis, the Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermohydraulical analysis of a Westinghouse 3-loop PWR plant with the MAAP code. The ISA methodology allows obtaining the SGTR Dynamic Event Tree taking into account the uncertainties on the operator actuation times. Simulations are performed with SCAIS (Simulation Code system for Integrated Safety Assessment), which includes a dynamic coupling with MAAP thermal hydraulic code. The offsite doses are calculated also with RADTRAD. The results shows the consequences of the sequences in terms not only of core damage but of offsite doses. This Thesis shows the need of analyzing all the consequences in an accident such as SGTR. For that, an it has been used an integral methodology like ISA-CSN. With this approach, the DSA vision of the SGTR (radiological consequences) is joined with the PSA vision of the SGTR (core damage consequences) to measure the total risk of the accident.

Neighborhood-corrected interface discontinuity factors for multi-group pin-by-pin diffusion calculations for LWR

Performing three-dimensional pin-by-pin full core calculations based on an improved solution of the multi-group diffusion equation is an affordable option nowadays to compute accurate local safety parameters for light water reactors. Since a transport approximation is solved, appropriate correction factors, such as interface discontinuity factors, are required to nearly reproduce the fully heterogeneous transport solution. Calculating exact pin-by-pin discontinuity factors requires the knowledge of the heterogeneous neutron flux distribution, which depends on the boundary conditions of the pin-cell as well as the local variables along the nuclear reactor operation. As a consequence, it is impractical to compute them for each possible configuration; however, inaccurate correction factors are one major source of error in core analysis when using multi-group diffusion theory. An alternative to generate accurate pin-by-pin interface discontinuity factors is to build a functional-fitting that allows incorporating the environment dependence in the computed values. This paper suggests a methodology to consider the neighborhood effect based on the Analytic Coarse-Mesh Finite Difference method for the multi-group diffusion equation. It has been applied to both definitions of interface discontinuity factors, the one based on the Generalized Equivalence Theory and the one based on Black-Box Homogenization, and for different few energy groups structures. Conclusions are drawn over the optimal functional-fitting and demonstrative results are obtained with the multi-group pin-by-pin diffusion code COBAYA3 for representative PWR configurations.

Analysis of the operator action and the single failure criteria in a SGTR sequence using best estimate assumptions with TRACE 5.0

Steam Generator Tube Rupture (SGTR) sequences in Pressurized Water Reactors are known to be one of the most demanding transients for the operating crew. SGTR are a special kind of transient as they could lead to radiological releases without core damage or containment failure, as they can constitute a direct path from the reactor coolant system to the environment. The first methodology used to perform the Deterministic Safety Analysis (DSA) of a SGTR did not credit the operator action for the first 30 min of the transient, assuming that the operating crew was able to stop the primary to secondary leakage within that period of time. However, the different real SGTR accident cases happened in the USA and over the world demonstrated that the operators usually take more than 30 min to stop the leakage in actual sequences. Some methodologies were raised to overcome that fact, considering operator actions from the beginning of the transient, as it is done in Probabilistic Safety Analysis. This paper presents the results of comparing different assumptions regarding the single failure criteria and the operator action taken from the most common methodologies included in the different Deterministic Safety Analysis. One single failure criteria that has not been analysed previously in the literature is proposed and analysed in this paper too. The comparison is done with a PWR Westinghouse three loop model in TRACE code (Almaraz NPP) with best estimate assumptions but including deterministic hypothesis such as single failure criteria or loss of offsite power. The behaviour of the reactor is quite diverse depending on the different assumptions made regarding the operator actions. On the other hand, although there are high conservatisms included in the hypothesis, as the single failure criteria, all the results are quite far from the regulatory limits. In addition, some improvements to the Emergency Operating Procedures to minimize the offsite release from the damaged SG in case of a SGTR are outlined taking into account the offsite dose sensitivity results.

Análisis de Ruido en Reactores PWR

Esta tesis se ha llevado a cabo persiguiendo dos objetivos principales: uno de ellos es el desarrollo y la aplicación de modelos para el mantenimiento predictivo de sensores en centrales nucleares, y el otro es profundizar en el entendimiento de los fenómenos que tienen influencia en el ruido de la señal de los detectores de neutrones de los reactores de agua a presión con ayuda de herramientas de simulación 3D. Para el desarrollo de los trabajos se ha contado con medidas de ruido de reactores PWR actualmente en operación registradas en el curso de la tesis. El análisis de estas medidas ha permitido desarrollar los modelos de los sensores a partir de sus señales reales y comparar lo obtenido en las simulaciones con la realidad. El estudio de los sensores y la elaboración de los modelos se han llevado a cabo mediante la aplicación de técnicas autorregresivas a las señales tomadas en planta. Para la reproducción de los fenómenos que tienen lugar en el núcleo del reactor y que pueden influir en el ruido neutrónico se ha contado con códigos neutrónicos ampliamente utilizados en la industria y con modelos actualizados y validados de las plantas. ABSTRACT There are two goals in this thesis. The first one is the development of models and its application for predictive maintenance of sensors in nuclear power plants. The second one is to improve the understanding of the phenomena that influence the neutron noise in pressurized water reactors by using 3D simulators. Real plant measurements recorded during this thesis have been used to achieve such goals. The information provided by the data led the development of the models and the comparison of the results provided by the computational simulations. Sensor models were obtained by applying autorregresive techniques to the signals recorded in the plant. Wide known codes in the nuclear industry as well as updated and validated models have been used for the reproduction of the phenomena that take place in the core an may influence the neutron noise.

Análisis de secuencias de accidente de pérdida de refrigerante en centrales nucleares de agua a presión mediante la metodología ISA

El accidente de pérdida de refrigerante (LOCA) en un reactor nuclear es uno de los accidentes Base de Diseño más preocupantes y estudiados desde el origen del uso de la tecnología de fisión en la industria productora de energía. El LOCA ocupa, desde el punto de vista de los análisis de seguridad, un lugar de vanguardia tanto en el análisis determinista (DSA) como probabilista (PSA), cuya diferenciada perspectiva ha ido evolucionando notablemente en lo que al crédito a la actuación de las salvaguardias y las acciones del operador se refiere. En la presente tesis se aborda el análisis sistemático de de las secuencias de LOCA por pequeña y mediana rotura en diferentes lugares de un reactor nuclear de agua a presión (PWR) con fallo total de Inyección de Seguridad de Alta Presión (HPSI). Tal análisis ha sido desarrollado en base a la metodología de Análisis Integrado de Seguridad (ISA), desarrollado por el Consejo de Seguridad Nuclear (CSN) y consistente en la aplicación de métodos avanzados de simulación y PSA para la obtención de Dominios de Daño, que cuantifican topológicamente las probabilidades de éxito y daño en función de determinados parámetros inciertos. Para la elaboración de la presente tesis, se ha hecho uso del código termohidráulico TRACE v5.0 (patch 2), avalado por la NRC de los EEUU como código de planta para la simulación y análisis de secuencias en reactores de agua ligera (LWR). Los objetivos del trabajo son, principalmente: (1) el análisis exhaustivo de las secuencias de LOCA por pequeña-mediana rotura en diferentes lugares de un PWR de tres lazos de diseño Westinghouse (CN Almaraz), con fallo de HPSI, en función de parámetros de gran importancia para los transitorios, tales como el tamaño de rotura y el tiempo de retraso en la respuesta del operador; (2) la obtención y análisis de los Dominios de Daño para transitorios de LOCA en PWRs, de acuerdo con la metodología ISA; y (3) la revisión de algunos de los resultados genéricos de los análisis de seguridad para secuencias de LOCA en las mencionadas condiciones. Los resultados de la tesis abarcan tres áreas bien diferenciadas a lo largo del trabajo: (a) la fenomenología física de las secuencias objeto de estudio; (b) las conclusiones de los análisis de seguridad practicados a los transitorios de LOCA; y (c) la relevancia de las consecuencias de las acciones humanas por parte del grupo de operación. Estos resultados, a su vez, son de dos tipos fundamentales: (1) de respaldo del conocimiento previo sobre el tipo de secuencias analizado, incluido en la extensa bibliografía examinada; y (2) hallazgos en cada una de las tres áreas mencionadas, no referidos en la bibliografía. En resumidas cuentas, los resultados de la tesis avalan el uso de la metodología ISA como método de análisis alternativo y sistemático para secuencias accidentales en LWRs. ABSTRACT The loss of coolant accident (LOCA) in nuclear reactors is one of the most concerning and analized accidents from the beginning of the use of fission technology for electric power production. From the point of view of safety analyses, LOCA holds a forefront place in both Deterministic (DSA) and Probabilistic Safety Analysis (PSA), which have significantly evolved from their original state in both safeguard performance credibility and human actuation. This thesis addresses a systematic analysis of small and medium LOCA sequences, in different places of a nuclear Pressurized Water Reactor (PWR) and with total failure of High Pressure Safety Injection (HPSI). Such an analysis has been grounded on the Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Regulatory Body (CSN). ISA involves the application of advanced methods of simulation and PSA for obtaining Damage Domains that topologically quantify the likelihood of success and damage regarding certain uncertain parameters.TRACE v5.0 (patch 2) code has been used as the thermalhydraulic simulation tool for the elaboration of this work. Nowadays, TRACE is supported by the US NRC as a plant code for the simulation and analysis of sequences in light water reactors (LWR). The main objectives of the work are the following ones: (1) the in-depth analysis of small and medium LOCA sequences in different places of a Westinghouse three-loop PWR (Almaraz NPP), with failed HPSI, regarding important parameters, such as break size or delay in operator response; (2) obtainment and analysis of Damage Domains related to LOCA transients in PWRs, according to ISA methodology; and (3) review some of the results of generic safety analyses for LOCA sequences in those conditions. The results of the thesis cover three separated areas: (a) the physical phenomenology of the sequences under study; (b) the conclusions of LOCA safety analyses; and (c) the importance of consequences of human actions by the operating crew. These results, in turn, are of two main types: (1) endorsement of previous knowledge about this kind of sequences, which is included in the literature; and (2) findings in each of the three aforementioned areas, not reported in the reviewed literature. In short, the results of this thesis support the use of ISA-like methodology as an alternative method for systematic analysis of LWR accidental sequences.

Reactor hot channel factors and burnout : a literature search /

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