NUCLEAR FISSION REACTORS

Author: ESCRIBANO MARTINEZ JOSE EDUARDO.
Year: 2003.
University: POLITÉCNICA DE MADRID [www.upm.es].
Place of defense: CENTRO DE LECTURA ESCUELA SUPERIOR INGENIEROS NAVALES.
Place of preparation: ESCUELA TECNICA SUPERIOR DE INGENIEROS NAVALES.

Summary: The aim of the thesis was to determine the criteria that define the Total Quality and methodologies for application to an industrial enterprise, which has resulted in a Modelo.También is the subject of the mismala measure of quality, as it affects ISO standards and national standards. In the first part, which includes capítuloos second to eleventh analyzes the proposed six criteria: management focused on the customer, strategic planning, leadership, human resources management processes and results in better continuing my implementation in the organization. Likewise it is considered a special quality in purchasing, production, inspection and also the costs of management including quality. On the other hand, and as an integral part of the process of implementing Total Quality and as a stepping stone to self discusses the Cetificación by Third Party Quality Management of an organization and / or its products and then consider the revision of Total Quality System by the Directorate, using the model developed in the thesis. The segund part is devoted to the study of the quality measures, aplicabales both control of procesoso commo their results, considrándose processes variability known and unknown latter circumstance requires the study of the vat of Studen off, whose knowledge is little widespread . Also considering the evolution of international standards ISO for inspection by variables that have replaced the traditional method of calculating development in the USA, for a graphical method of hard to justify.

Author: BATET MIRACLE LLUIS.
Year: 2003.
University: POLITÉCNICA DE CATALUÑA [www.upc.edu].
Place of defense: ETSEIB, DEPT. DE FÍSICA I ENG. NUCLEAR.
Place of preparation: EDIFICI B4 Campus NORD.

Summary: Forecasted primary energy consumption at the middle of XXI century will double or triple the present rate. Despite of the appeals for its reduction, annual emission of carbon dioxide to the atmosphere will continue rising. Electricity will cover an increasing part of the final energy demand. It seems plausible that nuclear power will continue playing an important role in the world energy supply. Nuclear power plants of new generation are designed adopting new safety approaches. Special consideration is deserved by those designs incorporating passive safety systems, which make use of natural forces (gravity, convection, etc.) to perform their function. This Ph.D. Thesis was originally conceived as a contribution to the development of passive safety systems. The cooperation established between the Technical University of Catalonia (UPC) and General Electric in the compass of the ESBWR project should have made it possible. Soon it was realized that the analytical tools used by the thermal-hydraulics community were not completely suitable to simulate accidental scenarios such as those anticipated for the passive type Advanced Light Water Reactors (ALWRs). During the late few years, the international community is working to establish the capabilities of the existing codes in relation to this concern. This Ph.D. Thesis aims to be a contribution in this direction. The main objective of this work is to demonstrate that, despite of their limitations, thermal-hydraulic system codes like RELAP5 are able to simulate, in a comprehensive approach, the behavior of an ALWR plant with passive safety features in accidental conditions, i.e. involving complex phenomena in the containment (mixing-stratification of vapor and noncondensable âNCâ gases, venting into the suppression pool), in the passive safety systems (condensation in presence of NC gases) and even in the reactor vessel (discharge of a gravity driven cooling system). To accomplish this objective, it has been essential the participation in two benchmark exercises based on experiments performed in the PANDA facility at the Paul Scherrer Institut, in Switzerland: the TEPSS project (funded by the European Community) and the ISP-42 exercise (under the auspices of the OECD) . One of the TEPSS objectives was the study of the residual heat removal by passive means after a Loss of Coolant Accident in the ESBWR. The tests also provided code assessment. ISP-42 interest was directly focused in testing the capabilities of present codes to deal with a new series of PANDA tests. RELAP5 code is used by the UPC to give support to plant operators and to the Nuclear Safety Council (CSN). The participation in TEPSS and ISP-42 permitted the UPC to test RELAP5 in challenging conditions. It was necessary to modify RELAP5 to improve its robustness in what concerns the steam condensation in presence of NC gases. Besides, the code normally does not permit the simulation of independent NC species, i.e. the composition of the gas mixture is the same in the whole system (it makes impossible, for example, to simulate the hydrogen release into the containment in case of a severe accident). One of the innovations of this work is that this limitation has been overcome. Of great importance is the co-operation agreement between the UPC and the Idaho National Engineering and Environmental Laboratory (INEEL) in order to use the RELAP5-3D code. This code includes three-dimensional (3D) components and it has been very helpful to corroborate, at the very end of this work, the rationale of different phenomena studied at the beginning using 1D options. It has been shown that RELAP5 is a robust code providing good comprehensive results when applied to the simulation of accidental scenarios of passive type ALWRs: it captures the general trends and the functionality of components and systems, imposing only a few boundary conditions. Thus, the code allows for global safety evaluations.

Author: VILLAMARÍN FERNÁNDEZ DAVID.
Year: 2004.
University: COMPLUTENSE DE MADRID [www.ucm.es].
Place of preparation: UNIVERSIDAD COMPLUTENSE DE MADRID.

Summary: The experiment MUSE-4 of 5Â ° Framework Program of the European Union has been carried out in the experimental reactor MASURCA located in Cadarache (France). This thesis focuses on the dynamic study of this whole spectrum subcritical fast coupled with a pulsed neutron source, as well as the desarro110, implementation and validation of experimental techniques to measure the reactivity not in need of a reference configuration in critical . The thesis has shown that by extending the model of the kinetic model to a spot several regions, it is entirely possible to describe the dynamics of the system with the parameters of the reactor (reactivity, steady and rapid decay lifetime average regions not multiplicativas reactor), except in the fifties microseconds immediately after the pulse neutron. For el1ose have conducted three pilot campaigns with reactividades ranging from keff = 0,995 to keff = 0.87, where it has been measured system response to a pulse of neutrons with a new data acquisition system developed by the author. Analysis of the experimental data was conducted using the method of areas Sjostrand and method of slope (extended model of kinetic spot). Once implemented corrections explained in the thesis has been found that both are compatible with each other, and which allow calculation of the reactivity with details better than 0.3% for Keff = 0.96. The methods of neutron noise, Rossi-alpha and Feynman-alpha show that very subcritical configurations is necessary to develop a new model to explain the experimental data. Also, the experimental data have been analyzed using the code Mó "Í1teCarIo MCNP, this has enabled explain in detail the results obtained, validating almismo time own code MCNP. Lastly, extrapolating the results obtained in the experiment MUSE-4 to designs ADS for transmutation, is studying the possible application of the techniques of measurement and determination of the reactivity thesis developed in the future transmutadores industry.

Author: ISERTE VILAR JOSÉ LUIS.
Year: 2005.
University: POLITÉCNICA DE VALENCIA [www.upv.es].
Place of defense: INGENIERIA QUIMICA Y NUCLEAR.
Place of preparation: UNIVERSIDAD POLITÉCNICA DE VALENCIA.

Summary: SUMMARY: This dissertation deals with the construction of new algorithms for the treatment of neutron signals. In particular, it seeks to obtain new tools algoritmicas to allow the characterization of the stability of nuclear reactors in boiling water. Specifically, it is possible to detect the oscillations of power these reactors. Sources of information on the system with which account is the signals from the monitoring system of power, which provides information on the neutron flux in the reactor core. Currently the basic parameters that characterize allow swings neutron flux in the reactor, as the reason buffer (RD), or the fundamental frequency, are generally assessed by methods based on the reconstruction of the signal through parametric autoregressive models (ARMA). These methods are working under the assumption of linear behavior of the reactor, but the behavior of nuclear reactors in boiling water can be approximated using only linear systems operating under normal conditions. In addition, these methods with some parameters depend on the order of the model chosen. This paper seeks methods to overcome these limitations, proposing algorithms whose application is not based on a model, ie without resorting to the use of assumptions simplificativas as the linear behavior of the system. Taking into account the above considerations, there are still two separate lines of research with mixed results. In the first are considered algorithms that allow reconstruction in the form of the analytic power spectral density (DEP), using the frequency moments. The previous approach was based on the problem of truncated Hamburger moments, and in this case we consider the problem truncation of Stieltjes, to allow use of the information provided by odd moments of the distribution function to rebuild. Besides the comprehensive information provided by the moment, we use the properties off of the signal using the local restrictions to transform Hilbert latter. In the second line of inquiry, considering that the signals under study is similar to a random process cuasi-estacionario for an interval of time sufficiently large, as is the case for signals from neutron reactor nuclear boiling water, it is estimated theoretically analitico behavior of the rules of Hilbert and Hilbert-Schmidt of matrices Toplitz formed from the spectral correlation function of this kind of signals. The asymptotic form of these rules is directly related to the oscillatory components not amortiguadas of random process. This is achieved by obtaining stability criteria for the signal, as well as a method for the detection of frequency instability. In order to study the reliability of the new methods, has been implementing them both simulated signals, which makes it possible to know a priori and accuracy characteristics of the signal under study, allowing a reliable comparison, as In this case, we have succeeded in creating a reliable algorithm detection and study of the instability of the reactors. The results obtained in the framework of the theory of moments with restrictions, to be more theoretical and the practical application of remote, even require some development to produce alternative algorithms in the 8 study 303 of behavior, stable and unstable nuclear reactors.

Author: CUERVO GOMEZ DIANA.
Year: 2006.
University: POLITÉCNICA DE MADRID [www.upm.es].
Place of defense: E.T.S. ING. INDUSTRIALES.
Place of preparation: E.T.S. ING. INDUSTRIALES.

Summary: The development of technology has enabled increasingly realistic view of the phenomena that occur in a nuclear reactor, and specifically in the areas neutron and termohidráulicos. This requires on the one hand the increase in the refinement of the computational mesh used, which has so far involved the homogenization of large regions of the reactor core, and on the other, the resolution of the form of the coupled equations. The nodalización which has been using three-dimensional calculations for the core is varied and generally different for calculating neutron and termohidráulico, which implies the use of average values of the conserved quantities in the interior of neutron node / channel cooling. In the field using neutron flux discontinuity factors for the correction of the average values that are derived from a preliminary estimate of two-dimensional diffusion in thin mesh. But this involves the use of correlations termohidráulicas for calculating the effective sections in addition to that, not being three-dimensional, it must use average values for the entire axial length. In the field termohidráulico has been the usual discretization of the reactor core in large regions where they are supposed homogeneous properties and despised therefore phenomena due to its internal geometry. The thesis has conducted a detailed study of the differences between a core calculation which uses a cooling channel for the fourth element of fuel, also called channel calculation means, and a detailed calculation of the cooling channel with representation from the subchannels that form. This would have analyzed the phenomena that occur within the fuel element and the effects of the use of average values in the nodes / channels account for the calculation of the nucleus. As part of this analysis have been obtained an equation for calculating the coefficient of enthalpy transport at the borders of the canal half from the values of enthalpy in the subchannels limit channel. This equation allows for the correction of equations termohidráulicas applied to the channel means to take into account the differences that occur in the calculation of gradients of the variables termohidráulicas in calculations of subchannels and channel medium. The analysis cited in the preceding paragraph has been applied to the nucleus of the NPP Ascó I realizándose a series of calculations in different situations variables reference power, flow and pressure in the full upper and at different times of the cycle, being used codes SIMTRAN and COBRA-IIIC/MIT-2 system SEANAP. In each of these situations has carried out a calculation of core complete and 68 calculations subchannels of the quarter-element comprising eighth-core, obtaining distributions differences between the two calculations. El uso del coeficiente de transporte de entalpía ha sido estudiado en un caso ejemplo habiendo supuesto una corrección de los valores de fracción de huecos a la salida del canal caliente en el cálculo de canal medio con respecto al cálculo detallado de un 4% y en general a redistribution of assets that approximates both calculations. Finally it has been implemented into the code COBRA-TF a new algorithm resolution matrix consisting of a Krylov method, which has shown an increase in the speed of the code up to 5 times the original speed, which will enable its use as part System SEANAP for calculating termohidráulico detailed in the scheme, local being developed.