Application of a Bayesian/Generalised Least-Squares method to generate correlations between independent neutron fission yield data

Fission product yields are fundamental parameters for several nuclear engineering calculations and in particular for burn-up/activation problems. The impact of their uncertainties was widely studied in the past and valuations were released, although still incomplete. Recently, the nuclear community expressed the need for full fission yield covariance matrices to produce inventory calculation results that take into account the complete uncertainty data. In this work, we studied and applied a Bayesian/generalised least-squares method for covariance generation, and compared the generated uncertainties to the original data stored in the JEFF-3.1.2 library. Then, we focused on the effect of fission yield covariance information on fission pulse decay heat results for thermal fission of 235U. Calculations were carried out using different codes (ACAB and ALEPH-2) after introducing the new covariance values. Results were compared with those obtained with the uncertainty data currently provided by the library. The uncertainty quantification was performed with the Monte Carlo sampling technique. Indeed, correlations between fission yields strongly affect the statistics of decay heat. Introduction Nowadays, any engineering calculation performed in the nuclear field should be accompanied by an uncertainty analysis. In such an analysis, different sources of uncertainties are taken into account. Works such as those performed under the UAM project (Ivanov, et al., 2013) treat nuclear data as a source of uncertainty, in particular cross-section data for which uncertainties given in the form of covariance matrices are already provided in the major nuclear data libraries. Meanwhile, fission yield uncertainties were often neglected or treated shallowly, because their effects were considered of second order compared to cross-sections (Garcia-Herranz, et al., 2010). However, the Working Party on International Nuclear Data Evaluation Co-operation (WPEC)

Processing and validation of JEFF3-3.1.2 Cross-section Library into Various Formats: ACE, PENDF, GENDF, MATXSR and BOXER.

Following the processing and validation of JEFF-3.1 performed in 2006 and presented in ND2007, and as a consequence of the latest updated of this library (JEFF-3.1.2) in February 2012, a new processing and validation of JEFF-3.1.2 cross section library is presented in this paper. The processed library in ACE format at ten different temperatures was generated with NJOY-99.364 nuclear data processing system. In addition, NJOY-99 inputs are provided to generate PENDF, GENDF, MATXSR and BOXER formats. The library has undergone strict QA procedures, being compared with other available libraries (e.g. ENDF/B-VII.1) and processing codes as PREPRO-2000 codes. A set of 119 criticality benchmark experiments taken from ICSBEP-2010 has been used for validation purposes.

Analysis of 238Pu and 56Fe Evaluated Data for Use in MYRRHA

A sensitivity analysis on the multiplication factor, keffkeff, to the cross section data has been carried out for the MYRRHA critical configuration in order to show the most relevant reactions. With these results, a further analysis on the 238Pu and 56Fe cross sections has been performed, comparing the evaluations provided in the JEFF-3.1.2 and ENDF/B-VII.1 libraries for these nuclides. Then, the effect in MYRRHA of the differences between evaluations are analysed, presenting the source of the differences. With these results, recommendations for the 56Fe and 238Pu evaluations are suggested. These calculations have been performed with SCALE6.1 and MCNPX-2.7e.

Propagation of Neutron Cross Section, Fission Yield, and Decay Data Uncertainties in Depletion Calculations

Propagation of nuclear data uncertainties in reactor calculations is interesting for design purposes and libraries evaluation. Previous versions of the GRS XSUSA library propagated only neutron cross section uncertainties. We have extended XSUSA uncertainty assessment capabilities by including propagation of fission yields and decay data uncertainties due to the their relevance in depletion simulations. We apply this extended methodology to the UAM6 PWR Pin-Cell Burnup Benchmark, which involves uncertainty propagation through burnup.

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.

Validation of the burn-up code EVOLCODE 2.0 with PWR experimental data and with a Sensitivity/Uncertainty analysis

A validation of the burn-up simulation system EVOLCODE 2.0 is presented here, involving the experimental measurement of U and Pu isotopes and some fission fragments production ratios after a burn-up of around 30 GWd/tU in a Pressurized Light Water Reactor (PWR). This work provides an in-depth analysis of the validation results, including the possible sources of the uncertainties. An uncertainty analysis based on the sensitivity methodology has been also performed, providing the uncertainties in the isotopic content propagated from the cross sections uncertainties. An improvement of the classical Sensitivity/ Uncertainty (S/U) model has been developed to take into account the implicit dependence of the neutron flux normalization, that is, the effect of the constant power of the reactor. The improved S/U methodology, neglected in this kind of studies, has proven to be an important contribution to the explanation of some simulation-experiment discrepancies for which, in general, the cross section uncertainties are, for the most relevant actinides, an important contributor to the simulation uncertainties, of the same order of magnitude and sometimes even larger than the experimental uncertainties and the experiment- simulation differences. Additionally, some hints for the improvement of the JEFF3.1.1 fission yield library and for the correction of some errata in the experimental data are presented.

Arquitectura antidesertización. Fluidez, biodiversidad, hidrofilia y transpirabilidad

La escasez de recursos, la desertización creciente y el previsible calentamiento global forman un escenario físico sin precedentes que urge la revisión de las relaciones entre arquitectura y agua en la urbanización de los paisajes secos, donde la aridez es el factor limitante principal. Cuestionar la idea de escasez, la de residuo o la de confort es el punto de partida para analizar la hidrología urbana. La condición ubicua y dinámica del agua, sus continuos cambios de estado y las implicaciones energéticas y ambientales involucradas argumentan a favor de un entendimiento integral, holístico, que aborda el diseño en relación a lo paisajístico, lo infraestructural y lo ambiental como un único ámbito de investigación, planteado en términos no solo de materia o de escasez, sino de energía. Este trabajo se interesa por el rol de la arquitectura en este proyecto conjunto, integral, del espacio físico con el ciclo hidrológico donde lo orgánico, lo geológico y lo atmosférico están ligados, y donde las categorías de lo hidrológico y lo hidráulico se confunden. Las profundas implicaciones culturales en la construcción de la naturaleza urbana y en la relación con el cuerpo humano adquieren especial notoriedad en los modelos de paisaje y de espacio público adoptados, importados desde las geografías húmedas, y en la estigmatización del agua desde los presupuestos del higienismo decimonónico, que conduce al hidrofugado general de una ciudad aséptica y estéril, que se extiende hasta las envolventes de fachadas estancas y los interiores blancos y satinados. La búsqueda de referentes salta la modernidad hacia contextos de baja energía como el oasis tradicional de las tierras áridas, que concilia la formación de un ecosistema productivo con la climatización de un espacio público exterior, o los ambientes aislados de la investigación aeroespacial, con ciclos cerrados de materia. Las condiciones del ciclo del agua, como la vinculación con el suelo y el territorio, la capacidad de disolver, de mezclarse y de lavar, su volumen variable y su papel como termorregulador señalan el interés que características físicas como la fluidez, la biodiversidad, la hidrofilia y la transpirabilidad tienen para una arquitectura con competencias hidráulicas y un espacio urbano transparente a los procesos del agua. La morfología y localización, la elección de escalas, jerarquías y relaciones entre espacios y la reformulación de los sistemas constructivos aparecen como herramientas y categorías propias desde las que proponer mejores respuestas a problemas como la deshidratación, la erosión y la contaminación. Una urbanización permeable e hidrófila que absorba el agua como un recurso valioso, nuevos ensamblajes para la bioquímica urbana (que introduzcan una idea de limpieza más próxima a fertilidad que a la desinfección), y el diseño de las condiciones atmosféricas a través de una arquitectura transpirable, que se empapa y rezuma frescor, son las claves de este nuevo proyecto. La ciudad se piensa como una síntesis multiescalar de espacios hidráulicos que aporta mayor resiliencia frente a la desertificación y las condiciones climáticas extremas, y mayor visibilidad en la escena pública al agua y a las inevitables conexiones entre ecología y economía. Pero también es una oportunidad para revisar las categorías disciplinares, para renovar las consistencias materiales, las calidades ambientales y las relaciones entre el cuerpo y el espacio. ABSTRACT The shortage of resources, foreseeable global warming and increasing desertification create an unprecedented prospect that question the existing relationships between water and architecture, in the urbanization of the arid lands. The awareness of the huge volumes of water that pierce unnoticed through urban space, their impact on the landscape as well as on environmental qualities, promote a design field where cultural, social and political considerations intersect, related to the body and the physical experience of space within the built environment. Fluidity and ubiquity, solution ability, variability and cyclical processes are characteristic of water as material, directly related with the fields of potential, chemical and thermal energy and the reality of its mass, as it occupies a changing volume in space. These are also the imposing cross sections that water introduces into the project, that argue in favor of a comprehensive and holistic understanding, of addressing design in relationship to landscape, infrastructure and environmental issues as a one single area of research. This work attempts to investigate how architecture, with its specific tools, can partake in the design of water cycle in the space, linking the organic, geological, and atmospheric, blurring the lines between hydrology and hydraulic. It aims to identify issues, within the continuous query associated with water, that deal with the architectural project and may have here better results. The deep cultural implications in the construction of urban nature and the relationship with the body, acquire special notoriety in the models of landscape and public space adopted, imported from humid geographies. Also in the stigmatization of water from the premises of nineteenth- century hygienics, which lead to the entire waterproofing of an aseptic and sterile city, to the sealed facades and white and polished interiors. The search for alternative references goes beyond modernity towards a mindset of low energy, as the traditional oasis of arid lands, which aims to reconcile the formation of a productive ecosystem with the conditioning of an outdoor public space, or the controlled environments of aerospace research, with closed cycles of matter. Fluidity, biodiversity, hydrophilicity and breathability are characteristic of an architecture with hydraulic competences. The distributing phenomenon of water, its necessary connectivity to the ground and to small cycles in the ecosystems, shows strong affinities with an infrastructural architecture, as an alternative to large-scale centralized networks. Its volume has approximated to the dimensions of the built space, promoting a new found condition of coexistence. A permeable and hydrophilic urbanization absorbs water as a valuable resource; new assemblies for urban biochemistry introduce an idea of sanitation closer to fertility than to disinfection; a breathable architecture that soaks and exudes freshness design the atmospheric conditions: these are the essential components of this new project. The city is understood as a synthesis of multi-scale hydraulic spaces that provides greater resilience against desertification and increases the visibility of water and the linkages between ecology and economy in the public scene. It is also an opportunity to review the disciplinary categories of architecture, the material consistencies, the environmental qualities and the relationship between body and space.

Effects of cross sections tables generation and optimization on rod ejection transient analyses

Best estimate analysis of rod ejection transients requires 3D kinetics core simulators. If they use cross sections libraries compiled in multidimensional tables,interpolation errors – originated when the core simulator computes the cross sections from the table values – are a source of uncertainty in k-effective calculations that should be accounted for. Those errors depend on the grid covering the domain of state variables and can be easily reduced, in contrast with other sources of uncertainties such as the ones due to nuclear data, by choosing an optimized grid distribution. The present paper assesses the impact of the grid structure on a PWR rod ejection transient analysis using the coupled neutron-kinetics/thermal-hydraulicsCOBAYA3/COBRA-TF system. Forthispurpose, the OECD/NEA PWR MOX/UO2 core transient benchmark has been chosen, as material compositions and geometries are available, allowing the use of lattice codes to generate libraries with different grid structures. Since a complete nodal cross-section library is also provided as part of the benchmark specifications, the effects of the library generation on transient behavior are also analyzed.Results showed large discrepancies when using the benchmark library and own-generated libraries when compared with benchmark participants’ solutions. The origin of the discrepancies was found to lie in the nodal cross sections provided in the benchmark.

Optimization of multidimensional cross-section tables for few-group core calculations

Multigroup diffusion codes for three dimensional LWR core analysis use as input data pre-generated homogenized few group cross sections and discontinuity factors for certain combinations of state variables, such as temperatures or densities. The simplest way of compiling those data are tabulated libraries, where a grid covering the domain of state variables is defined and the homogenized cross sections are computed at the grid points. Then, during the core calculation, an interpolation algorithm is used to compute the cross sections from the table values. Since interpolation errors depend on the distance between the grid points, a determined refinement of the mesh is required to reach a target accuracy, which could lead to large data storage volume and a large number of lattice transport calculations. In this paper, a simple and effective procedure to optimize the distribution of grid points for tabulated libraries is presented. Optimality is considered in the sense of building a non-uniform point distribution with the minimum number of grid points for each state variable satisfying a given target accuracy in k-effective. The procedure consists of determining the sensitivity coefficients of k-effective to cross sections using perturbation theory; and estimating the interpolation errors committed with different mesh steps for each state variable. These results allow evaluating the influence of interpolation errors of each cross section on k-effective for any combination of state variables, and estimating the optimal distance between grid points.

Limits to dark matter annihilation cross-section from a combined analysis of MAGIC and Fermi-LAT observations of dwarf satellite galaxies

We present the first joint analysis of gamma-ray data from the MAGIC Cherenkov telescopes and the Fermi Large Area Telescope (LAT) to search for gamma-ray signals from dark matter annihilation in dwarf satellite galaxies. We combine 158 hours of Segue 1 observations with MAGIC with 6-year observations of 15 dwarf satellite galaxies by the Fermi-LAT. We obtain limits on the annihilation cross-section for dark matter particle masses between 10 GeV and 100 TeV – the widest mass range ever explored by a single gamma-ray analysis. These limits improve on previously published Fermi-LAT and MAGIC results by up to a factor of two at certain masses. Our new inclusive analysis approach is completely generic and can be used to perform a global, sensitivity-optimized dark matter search by combining data from present and future gamma-ray and neutrino detectors.

Integral elastic, electronic-state, ionization, and total cross sections for electron scattering with furfural

We report absolute experimental integral cross sections (ICSs) for electron impact excitation of bands of electronic-states in furfural, for incident electron energies in the range 20-250 eV. Wherever possible, those results are compared to corresponding excitation cross sections in the structurally similar species furan, as previously reported by da Costa et al. [Phys. Rev. A 85, 062706 (2012)] and Regeta and Allan [Phys. Rev. A 91, 012707 (2015)]. Generally, very good agreement is found. In addition, ICSs calculated with our independent atom model (IAM) with screening corrected additivity rule (SCAR) formalism, extended to account for interference (I) terms that arise due to the multi-centre nature of the scattering problem, are also reported. The sum of those ICSs gives the IAM-SCAR+I total cross section for electron-furfural scattering. Where possible, those calculated IAM-SCAR+I ICS results are compared against corresponding results from the present measurements with an acceptable level of accord being obtained. Similarly, but only for the band I and band II excited electronic states, we also present results from our Schwinger multichannel method with pseudopotentials calculations. Those results are found to be in good qualitative accord with the present experimental ICSs. Finally, with a view to assembling a complete cross section data base for furfural, some binary-encounter-Bethe-level total ionization cross sections for this collision system are presented. (C) 2016 AIP Publishing LLC.

The Weakness of Post-Communist Civil Society Reassessed. CES Papers - Open Forum #11, 2012

During the last two decades, scholars from a variety of disciplines have argued that civil society is structurally deficient in post-communist countries. Yet why have the seemingly strong, active and mobilized civic movements of the transition period become so weak after democracy was established? And why have there been diverging political trajectories across the post-communist space if civil society structures were universally weak? This paper uses a wide range of data from various available sources to show that civil societies in Central and Eastern European countries are not as feeble as is commonly assumed. Some post-communist countries possess vigorous public spheres, and active civil society organizations strongly connected to transnational civic networks able to shape domestic policies. Following the calls by Anheier (2004) and Bernhard and Karakoç (2007) we adopt a multidimensional approach to the measurement of civil society. In a series of cross-section timeseries models, we show that our broader measures of civic and social institutions are able to predict the diverging transition paths among post-communist regimes, and in particular the growing gap between democratic East Central Europe and the increasingly authoritarian post-Soviet space.

Bibliography of low energy electron and photon cross section data (through December 1974) /

Includes index.

The integral Be⁹(n, 2n) cross section /

"MAy 14, 1964"--page 7.

Bibliography of low energy electron and photon cross section data, (January 1975 through December 1977) /

Includes indexes.