Anales de Edafología y Agrobiología Tomo 33 Número 9-10

Estudio del equilibrio nutritivo en cultivos de chirimoyo (Annona cherimolia), por César González O., Miguel Fuentes y Soledad Díaz.-- Resistencia a la desecación del tejido foliar y cierre de estomas en alfalfa (M edicago sativa L.) y trébol blanco (Trijolium repens L.) con relación al déficit agua, por M. Sánchez-Díaz y M. Sánchez-Marín.-- Determinación del calor isostérico y consideraciones sobre el mecanismo de la adsorción de fosfato por óxidos de hierro, por L. Madrid, F. Cabrera, P. de Arambarri y E. Díaz.-- Studies on sodium-calcium exchange equilibria. II. In Egyptian soils, by M. H. Nafady.-- Indices nutritivos en manzano (var. R. Delicious), por C. González, O. M. Rodríguez M., J. Solé D. y A. Wylie W.-- Caracteres de los suelos de las zonas citrícolas del valle de Murcia (España), por J. A. Sánchez F., F. Artes y J. López-Tarruella.-- Estudio micromorfológico de suelos desarrollados sobre andesitas en Andalucía oriental, por J. Aguilar y M. Delgado.-- Estudio edafológico de los relieves próximos a la vega de Motril, por J. Aguilar, Ruiz, A. Monge Ureña y C. Sierra Ruiz de la F.-- Consideraciones experimentales sobre el análisis de boro en plantas, por A. León, F.J. López-Andréu, F. Romojaro y C. Alcaraz.-- Efectos de la aplicación conjunta de fertilizantes químicos y microbianos (Azotobaeter Fosjobaeterias) en cultivos enarenados de tomate, por R. Azcón, M. Gómez y J. M. Barea.-- Formas de calcio en suelos del piso tropical de Barbacoas, Colombia, por G. Hugo Eraso, L. Federman Ortiz y O. Hernán Burbano.-- Compuestos íenólicos en Eriea vagans L., por J. Arinés, J. L. G. Mantilla y E. Vieitiz.-- Determinación de glúcidos en plantas por fotocolorimetría. Estudio comparativo de métodos clásicos y automáticos, por C. Cadahía y M. T. Piñeiro.-- Notas. Nombramiento de Consejeros Adjuntos del Patronato Alonso de Herrera.-- Nombramiento y cese de Vocales de la Junta de Gobierno del Patronato Alonso de Herrera.--Fallo de los Premios Alonso de Herrera y Antonio José de Cavanilles.-- Propuesta del Instituto de Alimentación y Productividad Animal sobre nombramiento de Vicedirector del mismo.-- Congresos y Reuniones internacionales.-- Creación de la Comisión Conjunta de Investigación Agraria de los Ministerios de Educación y Ciencia y de Agricultura.-- Restauración y adecuación del Jardín Botánico de Madrid.-- III Reunión Nacional de Centros de Investigación Ganadera Tribunales.-- Clausura del XI Curso Internacional de Edafología y Biología Vegetal.-- 7th International Colloquium on Plant Analysis and Fertilizer Problems.-- X Congreso Internacional de Ciencia del Suelo.-- Clausura del IV Congreso de Ciencia y Tecnología de Alimentos.-- 50th Anniversary Meeting of the British Society for Experimental Biología.-- XXIX Symposium on Symbiosis, Society for Experimental Biología.-- IV Reunión de la Sociedad Español de Microscopía Electrónica.-- Subvención de la Fundación Barrie de la Maza a la Misión Biológica de Galicia.-- Conferencia.-- Los universitarios y la defensa de la naturaleza.-- Seminario, sobre Tipos diferentes de costras calizas y su distribución regional.-- BibliografíaPeer reviewed2019-08.- CopyBook.- Libnova.- Biblioteca ICA

Nuclear data analyses for improving the safety of advanced lead-cooled reactors

El Reactor Rápido refrigerado por Plomo (LFR) es una de las tres tecnologías seleccionadas por la Plataforma Tecnológica de Energía Nuclear Sostenible que pueden satisfacer las futuras necesidades energéticas europeas. Investigadores e industria están realizando importantes esfuerzos para superar los principales inconvenientes del despliegue industrial de los LFR, que son la falta de experiencia operacional y el impacto de las incertidumbres en el diseño del reactor, la operación y la evaluación de la seguridad. En el diseño de reactores nucleares, las incertidumbres provienen principalmente de las propiedades de los materiales, las tolerancias de fabricación, las condiciones operativas, las herramientas de simulación y los datos nucleares. De hecho, la incertidumbre en los datos nucleares es una de las fuentes más importantes de incertidumbre en el diseño del reactor y en las simulaciones de la física del reactor y, en el pasado, se han obtenido sistemáticamente importantes diferencias entre las incertidumbres y las precisiones objetivo. Es necesario cumplir con la precisión objetivo no sólo para lograr el nivel de seguridad requerido para esta tecnología, sino también para minimizar el aumento de los costes debido a medidas de seguridad adicionales. Con esos antecedentes, el objetivo principal de este trabajo ha sido analizar y mejorar los datos nucleares necesarios para el desarrollo, la evaluación de seguridad y el licenciamiento de los reactores LFR, reduciendo las incertidumbres en los parámetros de reactividad (para seguridad) debido a las incertidumbres en los datos nucleares, con el fin de alcanzar las precisiones objetivo definidas por investigadores, industria y reguladores. Herramientas de sensibilidad e incertidumbre precisas y con alta fiabilidad son necesarias para estimar las incertidumbres en parámetros de seguridad clave del reactor (factor de multiplicación neutrónico, keff, fracción efectiva de neutrones diferidos, eff, tiempo efectivo de generación de neutrones, , coeficientes de reactividad, ...) e identificar posibles debilidades en los datos nucleares. Existen herramientas para calcular la incertidumbre de un parámetro del reactor debida a las incertidumbres en los datos nucleares. Sin embargo, estas herramientas poseen varias limitaciones, como carecer de capacidades de procesamiento en paralelo; necesidad de que el usuario seleccione los isótopos y canales de reacción a incluir en el análisis; uso de datos nucleares en estructura de multigrupos; uso de una librería de datos nucleares específica y/o una matriz de covarianza específica; y la limitación en la complejidad del sistema a analizar debido al número requerido de simulaciones. Por lo tanto, en este trabajo, se ha desarrollado una Metodología de Sensibilidad e Incertidumbre para códigos de MONtecarlo (SUMMON). SUMMON es una herramienta concebida para realizar análisis automatizados completos de sensibilidad e incertidumbre de los parámetros de reactividad (para seguridad) más relevantes de diseños de reactores desde el punto de vista neutrónico, es decir, keff, eff, eff y los coeficientes de reactividad, utilizando librerías de datos nucleares y covarianzas de última generación. SUMMON se ha validado utilizando experimentos integrales del ICSBEP (International Handbook of Evaluated Criticality Safety Benchmark Experiments) y se ha verificado exhaustivamente con códigos consolidados como SCALE, SUSD3D y SERPENT. Se ha encontrado un buen acuerdo entre los códigos. Una vez SUMMON fue desarrollado, se llevaron a cabo análisis preliminares para el diseño de MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications), un reactor rápido refrigerado con plomo-bismuto. Primero, se utilizó la librería de datos nucleares ENDF/B-VII.0 para identificar los datos nucleares más importantes para las reacciones inducidas por neutrones en los cálculos de criticidad de los LFR. Posteriormente, la librería JEFF-3.3T1, versión beta en ese momento de la nueva versión de la librería de datos nucleares evaluada en Europa, se analizó utilizando los mejores conjuntos de datos experimentales dependientes de la energía disponibles. El bismuto y el plomo, identificados en los análisis anteriores como isótopos clave, fueron seleccionados como los principales objetos de estudio para la mejora de los datos nucleares, ya que son de vital importancia y no fueron cubiertos en el proyecto piloto CIELO. Se encontraron problemas en la región de resonancias resueltas en las evaluaciones del plomo y el bismuto en JEFF-3.3T1 y se dieron recomendaciones al proyecto JEFF, que se adoptaron en la versión final de dicha librería de datos nucleares. A continuación, se realizaron análisis de sensibilidad e incertidumbre con las librerías de datos nucleares JEFF-3.3 y ENDF/B-VIII.0 mediante SUMMON para estimar las incertidumbres en los parámetros de criticidad de MYRRHA. Si bien se observó un buen acuerdo en las incertidumbres totales producidas por ambas librerías, las diferencias en las evaluaciones y la inexistencia de correlaciones y evaluaciones de covarianzas hicieron que los contribuyentes a la incertidumbre total difirieran. Además, las precisiones objetivo de diseño para algunos parámetros de seguridad, como el factor de multiplicación neutrónica, se excedieron en más del doble para las evaluaciones de datos nucleares consideradas. Con el fin de proporcionar datos nucleares ajustados, no sólo capaces de predecir las propiedades del reactor dentro de la precisión objetivo de diseño, sino también estadísticamente coherentes con los diversos experimentos diferenciales, se desarrolló el módulo de Asimilación de Datos Con summoN (DAWN). DAWN se basa en la combinación de datos de covarianza experimentales y experimentos integrales junto con técnicas avanzadas de ajuste estadístico (mínimos cuadrados generalizados). DAWN ha sido verificado utilizando el método TMC (Total Monte Carlo) para diferentes experimentos integrales. Finalmente, DAWN se usó para realizar una asimilación de los principales contribuyentes a la incertidumbre mediante el uso de datos nucleares de JEFF-3.3 a priori y experimentos de masa crítica disponibles públicamente en el ICSBEP. La consistencia del ajuste se verificó con datos experimentales diferenciales y se encontró un buen acuerdo. Se obtuvo una reducción significativa en la incertidumbre utilizando los experimentos más representativos de MYRRHA, debido a la reducción en la incertidumbre de los contribuyentes principales y la presencia a posteriori de fuertes correlaciones cruzadas entre isótopos y reacciones que no existían a priori. Los resultados muestran que se puede lograr una reducción de casi 300 pcm realizando una asimilación con el experimento más sensible al mayor contribuyente a la incertidumbre. Esto demuestra que la combinación de datos de covarianza experimental y experimentos integrales junto con la técnica de mínimos cuadrados generalizados puede proporcionar datos nucleares ajustados capaces de predecir las propiedades del reactor con menor incertidumbre, coherentes con los datos diferenciales. ----------ABSTRACT---------- The Lead-cooled Fast Reactor (LFR) is one of the three technologies selected by the Sustainable Nuclear Energy Technology Platform that can meet future European energy needs. Significant efforts are being made by researchers and industry to overcome the main drawbacks for the industrial deployment of LFR, which are the lack of operational experience and the impact of uncertainties in the reactor design, operation and safety assessment. In nuclear reactor design the uncertainties mainly come from material properties, fabrication tolerances, operative conditions, simulation tools and nuclear data. Indeed, the uncertainty in nuclear data is one of the most important sources of uncertainty in reactor design and reactor physics simulations, and significant gaps between the uncertainties and the target accuracies have been systematically shown in the past. Meeting the target accuracy is required not only to achieve the requested level of safety for this technology, but also to minimize the increase in the costs due to additional security measures. With that background, the main objective of this work has been to analyse and improve the nuclear data required for the development, safety assessment and licensing of LFR reactors, reducing the uncertainties in the criticality safety parameters due to the uncertainties in nuclear data, in order to reach the target accuracies defined by researchers, industry and regulators. To estimate the uncertainties in reactor key parameters (effective neutron multiplication factor, keff, effective delayed neutron fraction, βeff, effective neutron generation time, eff, safety coefficients, …) and to identify possible nuclear data weaknesses, accurate and reliable tools for sensitivity analysis and uncertainty quantification are needed. Tools able to calculate the uncertainty of a response due to uncertainties in nuclear data are available. However they possess several limitations such as no parallel processing capabilities; user selection of isotope and reaction channels to be included in the analysis; use of multi-group nuclear data; use of specific nuclear data library and/or specific covariance matrix; and limited complexity of the system under analysis due to the required number of simulations. Hence, in the framework of this work, a Sensitivity and Uncertainty Methodology for MONte carlo codes (SUMMON) has been developed. SUMMON is a tool conceived to perform complete automated sensitivity and uncertainty analyses of the most relevant criticality safety parameters of detailed complex reactor designs from the neutronic point of view, i.e., keff, eff, eff and reactivity coefficients, using state-of-the-art nuclear data libraries and covariances. SUMMON has been validated using integral experiments from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP) and extensively verified against consolidated codes such as SCALE, SUSD3D and SERPENT. Good agreement between codes has been found. Once SUMMON was developed, preliminary analyses were carried out for MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) lead-bismuth cooled fast reactor design. First, the ENDF/B-VII.0 nuclear data library was used, in order to identify the most important nuclear data for neutron induced reactions for criticality safety calculations of LFRs. Then, the recently released JEFF-3.3T1 library, the beta proposal at the time for the next version of the European evaluated nuclear data library, was analysed using the best documented energy dependent experimental data sets available. Bismuth and lead, identified in the previous analyses as key isotopes, were chosen as the main objects of study for improvement of nuclear data since they are of vital importance and were not covered in the CIELO pilot project. Problems were found in the resolved resonance region of JEFF-3.3T1 bismuth and lead evaluations and recommendations were given to the JEFF project, which were adopted in the release version of the library. Next, sensitivity and uncertainty analyses using the state-of-the-art JEFF-3.3 and ENDF/B-VIII.0 nuclear data libraries were performed with SUMMON to estimate the uncertainties in the criticality safety parameters of MYRRHA. While good agreement was observed in the total uncertainties yielded by both libraries, differences in evaluations, missing correlations and missing covariance evaluations, caused the contributors to the total uncertainty to differ. Furthermore, the design target accuracies for some criticality safety parameters, such as the effective neutron multiplication factor, still exceeded by more than a factor of two for the considered modern nuclear data evaluations. In order to provide adjusted nuclear data, not only capable of predicting reactor properties within the target design accuracy, but also statistically consistent with the various differential measurements, the Data Assimilation With summoN (DAWN) module was developed. DAWN is based on the combination of experimental covariance data and integral experiments together with advanced statistical adjustment techniques (Generalised Least Squares). DAWN has been verified against the Total Monte Carlo (TMC) method for several integral experiments. Finally, DAWN was used to perform an assimilation on the main contributors to the uncertainty using JEFF-3.3 nuclear data as a prior and publicly available critical mass experiments from the ICSBEP. The consistency of the nuclear data adjustment was checked against differential experimental data and good agreement was found. A significant reduction in uncertainty was obtained using the experiments most representative of MYRRHA, due to the reduction in the uncertainty of the major contributors and to the presence a posteriori of strong cross-correlations between isotopes and reactions that did not exist a priori. Results show that a reduction of nearly 300 pcm can be achieved performing an assimilation with the most sensitive experiment to the major contributor to the uncertainty. It proves that the combination of experimental covariance data and integral experiments together with Generalised Least Squares technique, can provide adjusted nuclear data capable of predicting reactor properties with lower uncertainty and consistent with differential data

Impact of nuclear data evaluations on data assimilation for an LFR

The Lead-cooled Fast Reactor is one of the three technologies selected by the Sustainable Nuclear Energy Technology Platform that can meet future European energy needs. The main drawbacks for the industrial deployment of LFR are the lack of operational experience and the impact of uncertainties. In nuclear reactor design the uncertainties mainly come from material properties, fabrication tolerances, operation conditions, simulation tools and nuclear data. The uncertainty in nuclear data is one of the most important sources of uncertainty in reactor physics simulations. Furthermore, it is known that the uncertainties in reactor criti-cality safety parameters are severely dependent on the nuclear data library used to estimate them. However, the impact of using different evaluations while performing data assimilation to constraint the uncertainties in the criticality parameters has not been properly assessed yet. In this work, a data assimilation for the main isotopes contributing to the uncertainty in keff of the ALFRED lead-cooled fast reactor has been performed with the SUMMON system using JEFF-3.3, ENDF/B-VIII.0 and JENDL-4.0u2 state-of-the-art nuclear data libraries, together with critical mass experiments from the International Criticality Safety Benchmark Evaluation Project that are representative of ALFRED, in order to assess the impact of using different evaluations for data assimilation

Impact of nuclear data evaluations on data assimilation for an LFR

The Lead-cooled Fast Reactor is one of the three technologies selected by the Sustainable Nuclear Energy Technology Platform that can meet future European energy needs. The main drawbacks for the industrial deployment of LFR are the lack of operational experience and the impact of uncertainties. In nuclear reactor design the uncertainties mainly come from material properties, fabrication tolerances, operation conditions, simulation tools and nuclear data. The uncertainty in nuclear data is one of the most important sources of uncertainty in reactor physics simulations. Furthermore, it is known that the uncertainties in reactor criti-cality safety parameters are severely dependent on the nuclear data library used to estimate them. However, the impact of using different evaluations while performing data assimilation to constraint the uncertainties in the criticality parameters has not been properly assessed yet. In this work, a data assimilation for the main isotopes contributing to the uncertainty in keff of the ALFRED lead-cooled fast reactor has been performed with the SUMMON system using JEFF-3.3, ENDF/B-VIII.0 and JENDL-4.0u2 state-of-the-art nuclear data libraries, together with critical mass experiments from the International Criticality Safety Benchmark Evaluation Project that are representative of ALFRED, in order to assess the impact of using different evaluations for data assimilation

Evolution of the importance of neutron-induced reactions along the cycle of an LFR

The Lead-cooled Fast Reactor is one of the three technologies selected by the Sustainable Nuclear Energy Technology Platform that can meet future European energy needs. Several LFR concepts are now in design phase, such as MYRRHA and ALFRED, and accurate nuclear data are required for the neutronic and safety assessment of the fast reactor designs. In this work, an assessment of the evolution of the importance of neutron-induced reactions along the cycle of a reference LFR design (i.e., ALFRED) with the state-of-the-art JEFF-3.3 nuclear data library is performed. Sensitivity analyses have been carried out with MCNP6 code in order to identify the most relevant isotopes and reactions from the neutronic point of view at BoL, BoC and EoC. Furthermore, an uncertainty quantification has been performed with the SUMMON system to study the evolution of uncertainties in the keff along the reactor cycle. The results from this work provide an exhaustive picture on the influence of nuclear data on core criticality performance, identifying key quantities and nuclear data needs relevant to achieve an improved safety level for LFR

Evolution of the importance of neutron-induced reactions along the cycle of an LFR

The Lead-cooled Fast Reactor is one of the three technologies selected by the Sustainable Nuclear Energy Technology Platform that can meet future European energy needs. Several LFR concepts are now in design phase, such as MYRRHA and ALFRED, and accurate nuclear data are required for the neutronic and safety assessment of the fast reactor designs. In this work, an assessment of the evolution of the importance of neutron-induced reactions along the cycle of a reference LFR design (i.e., ALFRED) with the state-of-the-art JEFF-3.3 nuclear data library is performed. Sensitivity analyses have been carried out with MCNP6 code in order to identify the most relevant isotopes and reactions from the neutronic point of view at BoL, BoC and EoC. Furthermore, an uncertainty quantification has been performed with the SUMMON system to study the evolution of uncertainties in the keff along the reactor cycle. The results from this work provide an exhaustive picture on the influence of nuclear data on core criticality performance, identifying key quantities and nuclear data needs relevant to achieve an improved safety level for LFR

Nuclear data analyses for improving the safety of advanced lead-cooled reactors

A target accuracy assessment of the effective neutron multiplication factor, keff, for MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) lead-bismuth cooled fast reactor has been performed with JEFF-3.3 and ENDF/B-VIII.0 state-of-the-art nuclear data libraries and the SUMMON system. Uncertainties in keff due to uncertainties in nuclear data have been assessed against the target accuracies provided by SG-26 of the WPEC of OECD/NEA in 2008 for LFR. Results show that keff target accuracy is still exceeded by more than a factor of two using the latest nuclear data evaluations released in 2018. Consequently, nuclear data assimilation has been carried out using criticality experiments from the International Criticality Safety Benchmark Evaluation Project that are representative of MYRRHA. The results from this work show that the level of accuracy needed in nuclear data cannot be obtained using only differential experiments, but the combination of experimental covariance data and integral experiments together with Generalised Least Squares technique can provide adjusted nuclear data capable of predicting reactor properties with lower uncertainty and consistent with differential data

Nuclear data analyses for improving the safety of advanced lead-cooled reactors

A target accuracy assessment of the effective neutron multiplication factor, keff, for MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) lead-bismuth cooled fast reactor has been performed with JEFF-3.3 and ENDF/B-VIII.0 state-of-the-art nuclear data libraries and the SUMMON system. Uncertainties in keff due to uncertainties in nuclear data have been assessed against the target accuracies provided by SG-26 of the WPEC of OECD/NEA in 2008 for LFR. Results show that keff target accuracy is still exceeded by more than a factor of two using the latest nuclear data evaluations released in 2018. Consequently, nuclear data assimilation has been carried out using criticality experiments from the International Criticality Safety Benchmark Evaluation Project that are representative of MYRRHA. The results from this work show that the level of accuracy needed in nuclear data cannot be obtained using only differential experiments, but the combination of experimental covariance data and integral experiments together with Generalised Least Squares technique can provide adjusted nuclear data capable of predicting reactor properties with lower uncertainty and consistent with differential data