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[EN] The neutron transport equation describes the distribution of neutrons inside a nuclear reactor core. Homogenization strategies have been used for decades to reduce the spatial and angular domain complexity of a nuclear reactor by replacing previously calculated heterogeneous subdomains by homogeneous ones and using a low order transport approximation to solve the new problem. The generalized equivalence theory for homogenization looks for discontinuous solutions through the introduction of discontinuity factors at the boundaries of the homogenized subdomains. In this work, the generalized equivalence theory is extended to the Simplified P-N equations using the finite element method. This extension proposes pin discontinuity factors instead of the usual assembly discontinuity factors and the use of the simplified spherical harmonics approximation rather than diffusion theory. An interior penalty finite element method is used to discretize and solve the problem using discontinuity factors. One dimensional numerical results show that the proposed pin discontinuity factors produce more accurate results than the usual assembly discontinuity factors. The proposed pin discontinuity factors produce precise results for both pin and assembly averaged values without using advanced reconstruction methods. Also, the homogenization methodology is verified against the calculation performed with reference discontinuity factors. (C) 2017 Elsevier B.V. All rights reserved., The work has been partially supported by the spanish Ministerio de Economía y Competitividad under projects ENE 2014-59442-P and MTM2014-58159-P, the Generalitat Valenciana under the project PROMETEO II/2014/008 and the Universitat Politècnica de València under the project FPI-2013. The work has also been supported partially by the Swedish Research Council (VR-Vetenskapsrådet) within a framework grant called DREAM4SAFER, research contract C0467701

In this work, BWR stability analysis has been performed on an operating point (PT_UPV) of Peach Bottom NPP which is inside the exclusion region in the operating power-flow map. The simulation has been made with the coupled code RELAP5-MOD3.3/PARCSv2.7. This point is achieved departing from test point 3 by a control rod movement as it is usually performed in Nuclear Power Plants. The thermal-hydraulic model is a detailed model that includes all the reactor vessel components: jet pumps, recirculation pumps, downcomer, reactor core and also the separator and the dryer. The reactor core has been modeled with 72 thermal-hydraulic channels, 71 represent the active core and 1 represents the core bypass. The reactor core thermal-hydraulic to neutronic representation (mapping) has been divided in four quadrants according to the first and second power harmonics (Lambda modes) obtained previously with the VALKIN code. This mapping was chosen in order not to condition the oscillation pattern. The transient starts with the control rod movement. The calculated results show that point PT_UPV is an unstable point and the obtained relative axial power distribution shows a bottom-peaked profile, which is characteristic of unstable cores.

En este trabajo se ha realizado un análisis de estabilidad tridimensional en el dominio temporal sobre la C.N. de Rhingals 1utilizando el código acoplado RELAPS5-MOD.3/PARCS v2.7.

En este trabajo, se ha realizado un análisis de estabilidad en los puntos de operación definidos en el bechmark de estabilidad de la central nuclear de Ringhals. Se quieren reproducir los resultados obtenidos en el bechmark con los códigos acoplados RELAP5-MOD3.3/PARCS v2.7 utilizando un modelo de reactor detallado. Las secciones eficaces, utilizadas por el código neutrónico, se han obtenido mediante la metodología SIMTAB desarrollada en la UPV junto con Iberdrola. El modelo termohidráulico incluye todos los componentes de la vasija del reactor: bombas de chorro, bombas de recirculación, downcomer, núcleo del reactor y también separadores y secadores de vapor. El equivalente entre el mallado termohidráulico y neutrónico (mapping) se ha dividido en cuatro cuadrantes según la forma del primer y segundo armónicos de la potencia (modos Lambda), obtenidos previamente con el código VALKIN. Con este procedimiento se pretende no condicionar el patrón de oscilación con la selección de canales termohidráulicos. El objetivo de este estudio es cualificar este código acoplado frente a este complejo tipo de accidentes tridimensionales que pueden producirse en el núcleo del reactor.

In a boiling water reactor (BWR), the power distribution can be varied both axially and radially by control rod manipulation and by void content variations in the moderator. As an example of accidental situations in nuclear power plants (NPP), we consider the power oscillations that can occur in a BWR NPP under certain circumstances. An interesting problem is to study and characterize these instabilities analyzing the neutronic power signals obtained from the Local Power Range Monitors (LPRMs) installed in the reactor core. Several techniques exist to detect and classify the possible oscillations in a BWR. The power modal decomposition method and the signal component analysis methodologies are reviewed. The performance of these techniques is compared analyzing Records 9 and 10 of cycle 14 of the OECD Ringhals 1 Stability Benchmark. © 2011 Elsevier Ltd. All rights reserved., This work has been partially supported by the Spanish Ministerio de Educacion y Ciencia under projects ENE2008-02669, the Generalitat Valenciana under Project ACOMP/2009/058, and the Universidad Politecnica de Valencia under project PAID-05-09-4285.