Iniciación a la energía nuclear : dos lecciones en la exposición sobre el átomo.

El hombre se ha preocupado siempre de conocer la estructura de la materia, del mundo que le rodea. El átomo de uranio sometido al bombardeo con neutrones sufre el fenómeno de la fisión. Una fisión del átomo es una rotura en dos trozos grandes.necesitamos una pareja de elementos que nos 92, puesto que el uranio tiene este número de cargas en su núcleo. Por ello, necesitamos tras la rotura dos elementos que sumen esa cantidad. Se conocen en la actualidad más de un centenar de formas diferentes de fisionarse el átomo de uranio, y no podemos predecir en un caso concreto como va a ser esa fisión. Pero si podemos explicar esa fisión a través del modelo que los físicos llaman el modelo nuclear de la gota litio. En el interior del núcleo las partículas están de la misma forma que las moléculas de una gota de un líquido. Estas moléculas son capaces de atraerse unas a otras en regiones muy pequeñas, cada molécula sólo es capaz de atraer a las moléculas que están cerca, no a las que están muy lejos. No todos los neutrones son eficaces en la rotura del átomo son más eficaces los neutrones lentos para ello, está el moderador, frenando los neutrones, haciéndoles chocar contra átomos de otros elementos, aumentaremos su eficacia y para eso está este aparato. A partir de estos elementos junto con una barra de control podremos fabricar un reactor, junto con refrigerantes y blindajes, es un proceso bastante complejo y no fácil de entender.

La exposición Átomos en Acción : más de 7.000 alumnos de Enseñanza Media, procedentes de toda España, asistieron a las clases de Introducción a la Ciencia nuclear. Aspectos científicos y pedagógicos.

Se realiza una valoración de la exposición Átomos en acción, que se clausuró el 15 de mayo de 1959 en Moncloa, Madrid. La exposición fue organizada por la Comisión de Energía Atómica de los Estados Unidos, en colaboración con la Junta de Energía Nuclear y el Centro de Orientación Didáctica de Enseñanza Media, ambas instituciones españolas. Se estima que unos 7000 alumnos de enseñanza media españoles acudieron a la exposición. En primer lugar se analiza como se explica la aplicación de la energía nuclear por medio de la exposición. La finalidad esencial de la exposición era ilustrar que en la fisión del átomo no debe asociarse siempre a las bombas atómicas. En segundo lugar se muestra como trabaja un reactor nuclear, como es su funcionamiento. Posteriormente se analizan los efectos de la fisión del átomo, que además de energía y calor, produce elementos como neutrones, rayos gamma e isótopos. Por otro lado se trataba de demostrar como hasta el cuerpo humano emite radioactividad, cuestión que fue de las que más curiosidad e impresión produjo a los visitantes a la exposición. También se señala que la exposición constituyó una demostración científica y pedagógica de primer orden, consiguiendo de una manera plena su finalidad primordialmente docente. Además la Exposición fue una empresa completamente altruista. Por último se hace mención a los tres objetivos que perseguía fundamentalmente: que el público en general se percatara de que la energía nuclear no es sólo la destructora bomba atómica; permitir a los científicos de la nación visitada el uso de un pequeño reactor nuclear, la instalación de un sistema de radiaciones gamma y otros aparatos y equipos en funcionamiento; y, por último, iniciar en la introducción de la ciencia atómica a los alumnos del Bachillerato y despertar en ellos la vocación a esta clase de conocimientos.

Sleep complaints and polysomnographic findings: A study of nuclear power plant shift workers

The literature widely recognizes that shift workers have more health complaints than the general population. The objective of this study was to describe the prevalence of sleep complaints and verify the polysomnographic (PSG) variables of shift workers in two Brazilian nuclear power plants. We carried out a subjective evaluation with a sleep questionnaire. Based on these results, the interviewees that reported sleep-related complaints were referred for polysomnographic evaluation. of the 327 volunteers initially evaluated by the sleep questionnaire, 113 (35%) reported sleep complaints; they were significantly older, had higher body mass index (BMI), and worked more years on shifts than those without sleep complaints. of these 113, 90 met criteria for various sleep disorders: 30 (9%) showed obstructive sleep apnea (OSA), 18 (5.5%) showed limb movement, and 42 (13%) evidenced both sleep problems and had a significantly higher proportion of sleep stage 1 and arousals compared with the 23 shift workers that had no indices of sleep problems. The present study found that 90 (27.5%) of the evaluated participants met the PSG criteria of some type of clinical sleep disorder. This high proportion should be investigated for associations with other aspects of work, such as working hours, working schedule, years performing shift work, and access to health services. Due to the strong association between sleep disorders and the incidence of fatigue and sleepiness, the evaluation of the sleep patterns and complaints of shift workers is essential and should be considered to be one of the basic strategies of industry to prevent accidents.

Contaminação radioativa: aspectos fundamentais associados à poluição nuclear

Since its discovery, radioactivity has brought numerous benefits to human societies. It has many applications in medicine, serving as a tool for non-invasive methods for diagnosis and therapies against diseases such as cancer. It also applies to technologies for energy in nuclear power plants with relatively low impacts on terms of perfect security. All applications, however, have risks, requiring maximum caution to drive processes and operations involving radioactive elements because, once released into the environment, they have extremely harmful effects on organisms affected. This paper presents fundamental concepts and principles of nuclear physics in order to understand the effects of radioactive elements released into the environment, culminating on the issue of radioactive contamination. Literature review allowed us to understand the radioactive contamination problem on living beings. Three major nuclear accidents have happened in the last thirty years, two of them in consecutive years. The nuclear accident at Chernobyl, Ukraine, in 1986, polluted large areas, condemning hundreds of thousands of people to live with consequences of the accident and effects of radiation, killing thousands of people throughout the years. In 1987, a major radiological accident occurred in Goiania (GO) when a source of radioactive cesium was violated, leading to the death of those who had direct or indirect contact with cesium. The most recent accident, in March, 2011, was located at the nuclear power plant in Fukushima Prefecture, Japan, after an earthquake and tsunami hit the region. There is no extensive and accurate knowledge about the consequences of the contamination entailed in that accident, although it is possible to verify signals on a global scale. An analysis of reports of contamination of large areas generated by nuclear plants with release of hazardous wastes suggests it is necessary to rethink the energy matrix of the various countries...

Bulk neutrinos as an alternative cause of the gallium and reactor anti-neutrino anomalies

We consider an alternative explanation for the deficit of nu(e) in Ga solar neutrino calibration experiments and of the (nu) over bar (e) in short-baseline reactor experiments by a model where neutrinos can oscillate into sterile Kaluza-Klein modes that can propagate in compactified submicrometer flat extra dimensions. We have analyzed the results of the gallium radioactive source experiments and 19 reactor experiments with baseline shorter than 100 m, and showed that these data can be fit into this scenario. The values of the lightest neutrino mass and of the size of the largest extra dimension that are compatible with these experiments are mostly not excluded by other neutrino oscillation experiments.

Power Transient Analysis of Experimental Devices for Jules Horowitz Material Testing Reactor (JHR)

The objective of this thesis is the power transient analysis concerning experimental devices placed within the reflector of Jules Horowitz Reactor (JHR). Since JHR material testing facility is designed to achieve 100 MW core thermal power, a large reflector hosts fissile material samples that are irradiated up to total relevant power of 3 MW. MADISON devices are expected to attain 130 kW, conversely ADELINE nominal power is of some 60 kW. In addition, MOLFI test samples are envisaged to reach 360 kW for what concerns LEU configuration and up to 650 kW according to HEU frame. Safety issues concern shutdown transients and need particular verifications about thermal power decreasing of these fissile samples with respect to core kinetics, as far as single device reactivity determination is concerned. Calculation model is conceived and applied in order to properly account for different nuclear heating processes and relative time-dependent features of device transients. An innovative methodology is carried out since flux shape modification during control rod insertions is investigated regarding the impact on device power through core-reflector coupling coefficients. In fact, previous methods considering only nominal core-reflector parameters are then improved. Moreover, delayed emissions effect is evaluated about spatial impact on devices of a diffuse in-core delayed neutron source. Delayed gammas transport related to fission products concentration is taken into account through evolution calculations of different fuel compositions in equilibrium cycle. Provided accurate device reactivity control, power transients are then computed for every sample according to envisaged shutdown procedures. Results obtained in this study are aimed at design feedback and reactor management optimization by JHR project team. Moreover, Safety Report is intended to utilize present analysis for improved device characterization.

The Effect of the Japan 2011 Disaster on Nuclear and Alternative Energy Stocks Worldwide: An Event Study

This event study investigates the impact of the Japanese nuclear disaster in Fukushima-Daiichi on the daily stock prices of French, German, Japanese, and U.S. nuclear utility and alternative energy firms. Hypotheses regarding the (cumulative) abnormal returns based on a three-factor model are analyzed through joint tests by multivariate regression models and bootstrapping. Our results show significant abnormal returns for Japanese nuclear utility firms during the one-week event window and the subsequent four-week post-event window. Furthermore, while French and German nuclear utility and alternative energy stocks exhibit significant abnormal returns during the event window, we cannot confirm abnormal returns for U.S. stocks.

Radiocarbon impact in a nearby tree of a light-water VVER type nuclear power plant, Paks, Hungary

Tree-ring series were collected for radiocarbon analyses from the vicinity of Paks nuclear power plant (NPP) and a background area (Dunaföldvár) for a 10-yr period (2000–2009). Samples of holocellulose were prepared from the wood and converted to graphite for accelerator mass spectrometry (AMS) 14C measurement using the MICADAS at ETH Zürich. The 14C concentration data from these tree rings was compared to the background tree rings for each year. The global decreasing trend of atmospheric 14C activity concentration was observed in the annual tree rings both in the background area and in the area of the NPP. As an average of the past 10 yr, the excess 14C emitted by the pressurized-water reactor (PWR) NPP to the atmosphere shows only a slight systematic excess (~6‰) 14C in the annual rings. The highest 14C excess was 13‰ (in 2006); however, years with the same 14C level as the background were quite frequent in the tree-ring series.

Propagation of Nuclear Data Uncertainties in Transmutation Calculations Using ACAB Code

The assessment of the accuracy of parameters related to the reactor core performance (e.g., ke) and f el cycle (e.g., isotopic evolution/transmutation) due to the uncertainties in the basic nuclear data (ND) is a critical issue. Different error propagation techniques (adjoint/forward sensitivity analysis procedures and/or Monte Carlo technique) can be used to address by computational simulation the systematic propagation of uncertainties on the final parameters. To perform this uncertainty assessment, the ENDF covariance les (variance/correlation in energy and cross- reactions-isotopes correlations) are required. In this paper, we assess the impact of ND uncertainties on the isotopic prediction for a conceptual design of a modular European Facility for Industrial Transmutation (EFIT) for a discharge burnup of 150 GWd/tHM. The complete set of uncertainty data for cross sections (EAF2007/UN, SCALE6.0/COVA-44G), radioactive decay and fission yield data (JEFF-3.1.1) are processed and used in ACAB code.

Isotopic prediction calculation methodologies: application to Vandellos-II Reactor cycles 7-11

Determining as accurate as possible spent nuclear fuel isotopic content is gaining importance due to its safety and economic implications. Since nowadays higher burn ups are achievable through increasing initial enrichments, more efficient burn up strategies within the reactor cores and the extension of the irradiation periods, establishing and improving computation methodologies is mandatory in order to carry out reliable criticality and isotopic prediction calculations. Several codes (WIMSD5, SERPENT 1.1.7, SCALE 6.0, MONTEBURNS 2.0 and MCNP-ACAB) and methodologies are tested here and compared to consolidated benchmarks (OECD/NEA pin cell moderated with light water) with the purpose of validating them and reviewing the state of the isotopic prediction capabilities. These preliminary comparisons will suggest what can be generally expected of these codes when applied to real problems. In the present paper, SCALE 6.0 and MONTEBURNS 2.0 are used to model the same reported geometries, material compositions and burn up history of the Spanish Van de llós II reactor cycles 7-11 and to reproduce measured isotopies after irradiation and decay times. We analyze comparisons between measurements and each code results for several grades of geometrical modelization detail, using different libraries and cross-section treatment methodologies. The power and flux normalization method implemented in MONTEBURNS 2.0 is discussed and a new normalization strategy is developed to deal with the selected and similar problems, further options are included to reproduce temperature distributions of the materials within the fuel assemblies and it is introduced a new code to automate series of simulations and manage material information between them. In order to have a realistic confidence level in the prediction of spent fuel isotopic content, we have estimated uncertainties using our MCNP-ACAB system. This depletion code, which combines the neutron transport code MCNP and the inventory code ACAB, propagates the uncertainties in the nuclide inventory assessing the potential impact of uncertainties in the basic nuclear data: cross-section, decay data and fission yields

Isotopic prediction simulations applied to high burnup samples irradiated in VANDELLÓS-II Reactor core

Isotopic content assessment has a paramount importance for safety and storage reasons. During the latest years, a great variety of codes have been developed to perform transport and decay calculations, but only those that couple both in an iterative manner achieve an accurate prediction of the final isotopic content of irradiated fuels. Needless to say, them all are supposed to pass the test of the comparison of their predictions against the corresponding experimental measures.

Temperature Evolution and Light Species Diffusion in Armor and Structural Material for Inertial Fusion Reactor Chambers: a Case for HiPER 4a

One of the most advance designs for HiPER fusion reactor is a spherical chamber 10 m in diameter based on dry wall concept. In this system, the first wall will have to withstand short energy pulses of 5 to 20 MJ at a repetition rate of 0.5-10 Hz mostly in form of X-rays and charged particles. To avoid melting of the inner surface, the first wall consists on a thin armor attached to the structural material. Thickness (th) and material of each layer have to be chosen to assure the proper functioning of the facility during its planned lifetime.

Silica final lenses under HiPER laser fusion reactor operation conditions

We have studied the thermo-mechanical response and atomistic degradation of final lenses in HiPER project. Final silica lenses are squares of 75 × 75 cm2 with a thickness of 5 cm. There are two scenarios where lenses are located at 8 m from the centre: •HiPER 4a, bunches of 100 shots (maximum 5 DT shots <48 MJ at ≈0.1 Hz). No blanket in chamber geometry. •HiPER 4b, continuous mode with shots ≈50 MJ at 10 Hz to generate 0.5 GW. Liquid metal blanket in chamber design.

Análisis de Estimación Óptima del Accidente de Pérdida de Refrigerante por Rotura Pequeña de 1”en Cabeza de Vasija en un Reactor tipo PWR

En el año 2002 durante una inspección se localizó una importante corrosión en la cabeza de la vasija de Davis Besse NPP. Si no se hubiera producido esa detección temprana, la corrosión hubiera provocado una pequeña rotura en la cabeza de la vasija. La OECD/NEA consideró la importancia de simular esta secuencia en la instalación experimental ROSA, la cual fue reproducida posteriormente por grupos de investigación internacionales con varios códigos de planta. En este caso el código utilizado para la simulación de las secuencias experimentales es TRACE. Los resultados de este test experimental fueron muy analizados internacionalmente por la gran influencia que dos factores tenía sobre el resultado: las acciones del operador relativas a la despresurización y la detección del descubrimiento del núcleo por los termopares que se encuentran a su salida. El comienzo del inicio de la despresurización del secundario estaba basado en la determinación del descubrimiento del núcleo por la lectura de los temopares de salida del núcleo. En el experimento se registró un retraso importante en la determinación de ese descubrimiento, comenzando la despresurización excesivamente tarde y haciendo necesaria la desactivación de los calentadores que simulan el núcleo del reactor para evitar su daño. Dada las condiciones excesivamente conservadoras del test experimentale, como el fallo de los dos trenes de inyección de alta presión durante todo el transitorio, en las aplicaciones de los experimentos con modelo de Almaraz NPP, se ha optado por reproducir dicho accidente con condiciones más realistas, verificando el impacto en los resultados de la disponibilidad de los trenes de inyección de alta presión o los tiempos de las acciones manuales del operador, como factores más limitantes y estableciendo el diámetro de rotura en 1”

Análisis de acciones del operador en un SBLOCA de 1”en el fondo de la Vasija en un Reactor tipo PWR

La simulación de accidentes de rotura pequeña en el fondo de la vasija se aparta del convencional análisis de LOCA de rama fría, el más limitante en los análisis deterministas La rotura de una de las penetraciones de instrumentación de la vasija ha sido desestimada históricamente en los análisis de licencia y en los Análisis Probabilistas de Seguridad y por ello, hay una falta evidente de literatura para dicho análisis. En el año 2003 durante una inspección, se detectó una considerable corrosión en el fondo de la vasija de South Texas Project Unit I NPP. La evolución en el tiempo de dicha corrosión habría derivado en una pequeña rotura en el fondo de la vasija si su detección no se hubiera producido a tiempo. La OECD/NEA consideró la importancia de simular dicha secuencia en la instalación experimental ROSA, la cual fue reproducida posteriormente por grupos de investigación internacionales con varios códigos de planta. En este caso el código utilizado para la simulación de las secuencias experimentales es TRACE. Tanto en el experimento como en la simulación se observaron las dificultades de reinundar la vasija al tener la rotura en el fondo de la misma, haciendo clave la gestión del accidente por parte del operador. Dadas las condiciones excesivamente conservadoras del test experimental, como el fallo de los dos trenes de inyección de alta presión durante todo el transitorio, en las aplicaciones de los experimentos con modelo de Almaraz NPP, se ha optado por reproducir dicho accidente con condiciones más realistas, verificando el impacto en los resultados de la disponibilidad de los trenes de inyección de alta presión o los tiempos de las acciones manuales del operador, como factores más limitantes y estableciendo el diámetro de rotura en 1”