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.

High power beam dump project for the accelerator prototype LIPAc: Cooling design and analysis

En el campo de la fusión nuclear y desarrollándose en paralelo a ITER (International Thermonuclear Experimental Reactor), el proyecto IFMIF (International Fusion Material Irradiation Facility) se enmarca dentro de las actividades complementarias encaminadas a solucionar las barreras tecnológicas que aún plantea la fusión. En concreto IFMIF es una instalación de irradiación cuya misión es caracterizar materiales resistentes a condiciones extremas como las esperadas en los futuros reactores de fusión como DEMO (DEMOnstration power plant). Consiste de dos aceleradores de deuterones que proporcionan un haz de 125 mA y 40 MeV cada uno, que al colisionar con un blanco de litio producen un flujo neutrónico intenso (1017 neutrones/s) con un espectro similar al de los neutrones de fusión [1], [2]. Dicho flujo neutrónico es empleado para irradiar los diferentes materiales candidatos a ser empleados en reactores de fusión, y las muestras son posteriormente examinadas en la llamada instalación de post-irradiación. Como primer paso en tan ambicioso proyecto, una fase de validación y diseño llamada IFMIFEVEDA (Engineering Validation and Engineering Design Activities) se encuentra actualmente en desarrollo. Una de las actividades contempladas en esta fase es la construcción y operación de una acelarador prototipo llamado LIPAc (Linear IFMIF Prototype Accelerator). Se trata de un acelerador de deuterones de alta intensidad idéntico a la parte de baja energía de los aceleradores de IFMIF. Los componentes del LIPAc, que será instalado en Japón, son suministrados por diferentes países europeos. El acelerador proporcionará un haz continuo de deuterones de 9 MeV con una potencia de 1.125 MW que tras ser caracterizado con diversos instrumentos deberá pararse de forma segura. Para ello se requiere un sistema denominado bloque de parada (Beam Dump en inglés) que absorba la energía del haz y la transfiera a un sumidero de calor. España tiene el compromiso de suministrar este componente y CIEMAT (Centro de Investigaciones Energéticas Medioambientales y Tecnológicas) es responsable de dicha tarea. La pieza central del bloque de parada, donde se para el haz de iones, es un cono de cobre con un ángulo de 3.5o, 2.5 m de longitud y 5 mm de espesor. Dicha pieza está refrigerada por agua que fluye en su superficie externa por el canal que se forma entre el cono de cobre y otra pieza concéntrica con éste. Este es el marco en que se desarrolla la presente tesis, cuyo objeto es el diseño del sistema de refrigeración del bloque de parada del LIPAc. El diseño se ha realizado utilizando un modelo simplificado unidimensional. Se han obtenido los parámetros del agua (presión, caudal, pérdida de carga) y la geometría requerida en el canal de refrigeración (anchura, rugosidad) para garantizar la correcta refrigeración del bloque de parada. Se ha comprobado que el diseño permite variaciones del haz respecto a la situación nominal siendo el flujo crítico calorífico al menos 2 veces superior al nominal. Se han realizado asimismo simulaciones fluidodinámicas 3D con ANSYS-CFX en aquellas zonas del canal de refrigeración que lo requieren. El bloque de parada se activará como consecuencia de la interacción del haz de partículas lo que impide cualquier cambio o reparación una vez comenzada la operación del acelerador. Por ello el diseño ha de ser muy robusto y todas las hipótesis utilizadas en la realización de éste deben ser cuidadosamente comprobadas. Gran parte del esfuerzo de la tesis se centra en la estimación del coeficiente de transferencia de calor que es determinante en los resultados obtenidos, y que se emplea además como condición de contorno en los cálculos mecánicos. Para ello por un lado se han buscado correlaciones cuyo rango de aplicabilidad sea adecuado para las condiciones del bloque de parada (canal anular, diferencias de temperatura agua-pared de decenas de grados). En un segundo paso se han comparado los coeficientes de película obtenidos a partir de la correlación seleccionada (Petukhov-Gnielinski) con los que se deducen de simulaciones fluidodinámicas, obteniendo resultados satisfactorios. Por último se ha realizado una validación experimental utilizando un prototipo y un circuito hidráulico que proporciona un flujo de agua con los parámetros requeridos en el bloque de parada. Tras varios intentos y mejoras en el experimento se han obtenido los coeficientes de película para distintos caudales y potencias de calentamiento. Teniendo en cuenta la incertidumbre de las medidas, los valores experimentales concuerdan razonablemente bien (en el rango de 15%) con los deducidos de las correlaciones. Por motivos radiológicos es necesario controlar la calidad del agua de refrigeración y minimizar la corrosión del cobre. Tras un estudio bibliográfico se identificaron los parámetros del agua más adecuados (conductividad, pH y concentración de oxígeno disuelto). Como parte de la tesis se ha realizado asimismo un estudio de la corrosión del circuito de refrigeración del bloque de parada con el doble fin de determinar si puede poner en riesgo la integridad del componente, y de obtener una estimación de la velocidad de corrosión para dimensionar el sistema de purificación del agua. Se ha utilizado el código TRACT (TRansport and ACTivation code) adaptándalo al caso del bloque de parada, para lo cual se trabajó con el responsable (Panos Karditsas) del código en Culham (UKAEA). Los resultados confirman que la corrosión del cobre en las condiciones seleccionadas no supone un problema. La Tesis se encuentra estructurada de la siguiente manera: En el primer capítulo se realiza una introducción de los proyectos IFMIF y LIPAc dentro de los cuales se enmarca esta Tesis. Además se describe el bloque de parada, siendo el diseño del sistema de rerigeración de éste el principal objetivo de la Tesis. En el segundo y tercer capítulo se realiza un resumen de la base teórica así como de las diferentes herramientas empleadas en el diseño del sistema de refrigeración. El capítulo cuarto presenta los resultados del relativos al sistema de refrigeración. Tanto los obtenidos del estudio unidimensional, como los obtenidos de las simulaciones fluidodinámicas 3D mediante el empleo del código ANSYS-CFX. En el quinto capítulo se presentan los resultados referentes al análisis de corrosión del circuito de refrigeración del bloque de parada. El capítulo seis se centra en la descripción del montaje experimental para la obtención de los valores de pérdida de carga y coeficiente de transferencia del calor. Asimismo se presentan los resultados obtenidos en dichos experimentos. Finalmente encontramos un capítulo de apéndices en el que se describen una serie de experimentos llevados a cabo como pasos intermedios en la obtención del resultado experimental del coeficiente de película. También se presenta el código informático empleado para el análisis unidimensional del sistema de refrigeración del bloque de parada llamado CHICA (Cooling and Heating Interaction and Corrosion Analysis). ABSTRACT In the nuclear fusion field running in parallel to ITER (International Thermonuclear Experimental Reactor) as one of the complementary activities headed towards solving the technological barriers, IFMIF (International Fusion Material Irradiation Facility) project aims to provide an irradiation facility to qualify advanced materials resistant to extreme conditions like the ones expected in future fusion reactors like DEMO (DEMOnstration Power Plant). IFMIF consists of two constant wave deuteron accelerators delivering a 125 mA and 40 MeV beam each that will collide on a lithium target producing an intense neutron fluence (1017 neutrons/s) with a similar spectra to that of fusion neutrons [1], [2]. This neutron flux is employed to irradiate the different material candidates to be employed in the future fusion reactors, and the samples examined after irradiation at the so called post-irradiative facilities. As a first step in such an ambitious project, an engineering validation and engineering design activity phase called IFMIF-EVEDA (Engineering Validation and Engineering Design Activities) is presently going on. One of the activities consists on the construction and operation of an accelerator prototype named LIPAc (Linear IFMIF Prototype Accelerator). It is a high intensity deuteron accelerator identical to the low energy part of the IFMIF accelerators. The LIPAc components, which will be installed in Japan, are delivered by different european countries. The accelerator supplies a 9 MeV constant wave beam of deuterons with a power of 1.125 MW, which after being characterized by different instruments has to be stopped safely. For such task a beam dump to absorb the beam energy and take it to a heat sink is needed. Spain has the compromise of delivering such device and CIEMAT (Centro de Investigaciones Energéticas Medioambientales y Tecnológicas) is responsible for such task. The central piece of the beam dump, where the ion beam is stopped, is a copper cone with an angle of 3.5o, 2.5 m long and 5 mm width. This part is cooled by water flowing on its external surface through the channel formed between the copper cone and a concentric piece with the latter. The thesis is developed in this realm, and its objective is designing the LIPAc beam dump cooling system. The design has been performed employing a simplified one dimensional model. The water parameters (pressure, flow, pressure loss) and the required annular channel geometry (width, rugoisty) have been obtained guaranteeing the correct cooling of the beam dump. It has been checked that the cooling design allows variations of the the beam with respect to the nominal position, being the CHF (Critical Heat Flux) at least twice times higher than the nominal deposited heat flux. 3D fluid dynamic simulations employing ANSYS-CFX code in the beam dump cooling channel sections which require a more thorough study have also been performed. The beam dump will activateasaconsequenceofthe deuteron beam interaction, making impossible any change or maintenance task once the accelerator operation has started. Hence the design has to be very robust and all the hypotheses employed in the design mustbecarefully checked. Most of the work in the thesis is concentrated in estimating the heat transfer coefficient which is decisive in the obtained results, and is also employed as boundary condition in the mechanical analysis. For such task, correlations which applicability range is the adequate for the beam dump conditions (annular channel, water-surface temperature differences of tens of degrees) have been compiled. In a second step the heat transfer coefficients obtained from the selected correlation (Petukhov- Gnielinski) have been compared with the ones deduced from the 3D fluid dynamic simulations, obtaining satisfactory results. Finally an experimental validation has been performed employing a prototype and a hydraulic circuit that supplies a flow with the requested parameters in the beam dump. After several tries and improvements in the experiment, the heat transfer coefficients for different flows and heating powers have been obtained. Considering the uncertainty in the measurements the experimental values agree reasonably well (in the order of 15%) with the ones obtained from the correlations. Due to radiological reasons the quality of the cooling water must be controlled, hence minimizing the copper corrosion. After performing a bibligraphic study the most adequate water parameters were identified (conductivity, pH and dissolved oxygen concentration). As part of this thesis a corrosion study of the beam dump cooling circuit has been performed with the double aim of determining if corrosion can pose a risk for the copper beam dump , and obtaining an estimation of the corrosion velocitytodimension the water purification system. TRACT code(TRansport and ACTivation) has been employed for such study adapting the code for the beam dump case. For such study a collaboration with the code responsible (Panos Karditsas) at Culham (UKAEA) was established. The work developed in this thesis has supposed the publication of three articles in JCR journals (”Journal of Nuclear Materials” y ”Fusion Engineering and Design”), as well as presentations in more than four conferences and relevant meetings.

Report on containment studies of the Enrico Fermi Atomic Power Plant /

Includes bibliographies.

Light water breeder reactor program.

"June 1976."

Evaluation of radioiodine measurements at Pilgrim Nuclear Power Plant /

Manuscript completed September 1978, published October 1978.

A three region steam drum model for a nuclear power plant simulator (BRENDA) /

"Date published: August 1981."

Fast breeder reactor studies /

"Under contract W-31-109-Eng-38."

Status report[s] on Clinch River Study.

UC-70

Active and reactive power control of synchronous generator for the realization of a virtual power plant

This paper presents a novel power control strategy that decouples the active and reactive power for a synchronous generator connected to a power network. The proposed control paradigm considers the capacitance of the transmission line along with its resistance and reactance as-well. Moreover the proposed controller takes into account all cases of R-X relationships, thus allowing it to function in Virtual Power Plant (VPP) structures which operate at both medium voltage (MV) and low voltage (LV) levels. The independent control of active and reactive power is achieved through rotational transformations of the terminal voltages and currents at the synchronous generator's output. This paper details the control technique by first presenting the mathematical and electrical network analysis of the methodology and then successfully implementing the control using MATLAB-SIMULINK simulation.

The Fukushima nuclear power plant disaster and percetions of health risk communication : a case study

This case study was conducted to explore the perceptions of health risk messages sent by the Japanese Government following the Fukushima nuclear power plant disaster. The content of health risk messages from the Japanese Government and the Japanese national broadcaster (NHK) were analysed and semi-structured interviews were conducted with a sample of Tokyo residents. Initially, participants trusted these messages but as the crisis unfolded they became sceptical about the messages. Participants felt the messages did not communicate health risk information effectively because the messages were; not supported by evidence, inconsistent, delayed and changed over time. Despite widespread access to the internet, social media and mobile telephones, most participants relied on television news for information about the health risks. The Japanese Government urgently needs to re-build trust by engaging the community in the planning and development phases of health risk communication strategies.

Assessing organizational culture in complex sociotechnical systems : Methodological evidence from studies in nuclear power plant maintenance organizations

Failures in industrial organizations dealing with hazardous technologies can have widespread consequences for the safety of the workers and the general population. Psychology can have a major role in contributing to the safe and reliable operation of these technologies. Most current models of safety management in complex sociotechnical systems such as nuclear power plant maintenance are either non-contextual or based on an overly-rational image of an organization. Thus, they fail to grasp either the actual requirements of the work or the socially-constructed nature of the work in question. The general aim of the present study is to develop and test a methodology for contextual assessment of organizational culture in complex sociotechnical systems. This is done by demonstrating the findings that the application of the emerging methodology produces in the domain of maintenance of a nuclear power plant (NPP). The concepts of organizational culture and organizational core task (OCT) are operationalized and tested in the case studies. We argue that when the complexity of the work, technology and social environment is increased, the significance of the most implicit features of organizational culture as a means of coordinating the work and achieving safety and effectiveness of the activities also increases. For this reason a cultural perspective could provide additional insight into the problem of safety management. The present study aims to determine; (1) the elements of the organizational culture in complex sociotechnical systems; (2) the demands the maintenance task sets for the organizational culture; (3) how the current organizational culture at the case organizations supports the perception and fulfilment of the demands of the maintenance work; (4) the similarities and differences between the maintenance cultures at the case organizations, and (5) the necessary assessment of the organizational culture in complex sociotechnical systems. Three in-depth case studies were carried out at the maintenance units of three Nordic NPPs. The case studies employed an iterative and multimethod research strategy. The following methods were used: interviews, CULTURE-survey, seminars, document analysis and group work. Both cultural analysis and task modelling were carried out. The results indicate that organizational culture in complex sociotechnical systems can be characterised according to three qualitatively different elements: structure, internal integration and conceptions. All three of these elements of culture as well as their interrelations have to be considered in organizational assessments or important aspects of the organizational dynamics will be overlooked. On the basis of OCT modelling, the maintenance core task was defined as balancing between three critical demands: anticipating the condition of the plant and conducting preventive maintenance accordingly, reacting to unexpected technical faults and monitoring and reflecting on the effects of maintenance actions and the condition of the plant. The results indicate that safety was highly valued at all three plants, and in that sense they all had strong safety cultures. In other respects the cultural features were quite different, and thus the culturally-accepted means of maintaining high safety also differed. The handicraft nature of maintenance work was emphasised as a source of identity at the NPPs. Overall, the importance of safety was taken for granted, but the cultural norms concerning the appropriate means to guarantee it were little reflected. A sense of control, personal responsibility and organizational changes emerged as challenging issues at all the plants. The study shows that in complex sociotechnical systems it is both necessary and possible to analyse the safety and effectiveness of the organizational culture. Safety in complex sociotechnical systems cannot be understood or managed without understanding the demands of the organizational core task and managing the dynamics between the three elements of the organizational culture.

Design of a Near Optimum Regulator for a Power Plant

For systems which can be decomposed into slow and fast subsystems, a near optimum linear state regulator consisting of two subsystem regulators can be developed. Depending upon the desired criteria, either a short term (fast controller) or a long term controller (slow controller) can be easily designed with minimum computational costs. Using this approach an example of a power system supplying a cyclic load is studied and the performance of the different controllers are compared.

Simulation of Dust Loaded V-I Characteristics of a Commercial Thermal Power Plant Precipitator

A new approach based on finite difference method, is proposed for the simulation of electrical conditions in a dc energized wire-duct electrostatic precipitator with and without dust loading. Simulated voltage-curren characteristics with and without dust loading were compared with the measured characteristics for analyzing the performance of a precipitator. The simple finite difference method gives sufficiently accurate results with reduced mesh size. The results for dust free simulation were validated with published experimental data. Further measurements were conducted at a thermal power plant in India and the results compares well with the measured ones.

Seismic microzonation of a nuclear power plant site with detailed geotechnical, geophysical and site effect studies

This paper highlights the seismic microzonation carried out for a nuclear power plant site. Nuclear power plants are considered to be one of the most important and critical structures designed to withstand all natural disasters. Seismic microzonation is a process of demarcating a region into individual areas having different levels of various seismic hazards. This will help in identifying regions having high seismic hazard which is vital for engineering design and land-use planning. The main objective of this paper is to carry out the seismic microzonation of a nuclear power plant site situated in the east coast of South India, based on the spatial distribution of the hazard index value. The hazard index represents the consolidated effect of all major earthquake hazards and hazard influencing parameters. The present work will provide new directions for assessing the seismic hazards of new power plant sites in the country. Major seismic hazards considered for the evaluation of the hazard index are (1) intensity of ground shaking at bedrock, (2) site amplification, (3) liquefaction potential and (4) the predominant frequency of the earthquake motion at the surface. The intensity of ground shaking in terms of peak horizontal acceleration (PHA) was estimated for the study area using both deterministic and probabilistic approaches with logic tree methodology. The site characterization of the study area has been carried out using the multichannel analysis of surface waves test and available borehole data. One-dimensional ground response analysis was carried out at major locations within the study area for evaluating PHA and spectral accelerations at the ground surface. Based on the standard penetration test data, deterministic as well as probabilistic liquefaction hazard analysis has been carried out for the entire study area. Finally, all the major earthquake hazards estimated above, and other significant parameters representing local geology were integrated using the analytic hierarchy process and hazard index map for the study area was prepared. Maps showing the spatial variation of seismic hazards (intensity of ground shaking, liquefaction potential and predominant frequency) and hazard index are presented in this work.