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Espinosa-Paredes G.,Metropolitan Autonomous University | Polo-Labarrios M.-A.,Comision Nacional de Seguridad Nuclear y Salvaguardias
Annals of Nuclear Energy | Year: 2016

The aim of this work is investigate the effects of each term of fractional neutron point kinetics (FNPK) equations. The FNPK approach includes three fractional derivatives, which are explored in this work to understand their effect on the neutron density, the concentration of neutron precursors, and the source term. According with the results of the numerical experiments, the derivative with respect to time of the source term is negligible and the FNPK is simplified. © 2016 Elsevier Ltd. All rights reserved.

Nunez-Carrera A.,Comision Nacional de Seguridad Nuclear y Salvaguardias | Camargo-Camargo R.,Comision Nacional de Seguridad Nuclear y Salvaguardias | Espinosa-Paredes G.,Metropolitan Autonomous University | Lopez-Garcia A.,Comision Nacional de Seguridad Nuclear y Salvaguardias
Science and Technology of Nuclear Installations | Year: 2012

The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR) lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV). The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA) with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation. Copyright © 2012 Alejandro Nuez-Carrera et al.

Nunez-Carrera A.,Comision Nacional de Seguridad Nuclear y Salvaguardias | Espinosa-Paredes G.,Metropolitan Autonomous University | Cruz-Esteban H.,National Polytechnic Institute of Mexico
Nuclear Engineering and Design | Year: 2011

A fuzzy cognitive maps (FCM) application is proposed as a simple method to determine failure modes and effects of the standby liquid control system (SLC) during anticipated transient without scram (ATWS) in a boiling water reactor (BWR). The SLC has an important contribution to the total core damage frequency in a BWR. This is the first step in the development of an expert system that could involve many emergency systems of a BWR to simulate accident sequences, through the knowledge representation and reasoning with FCM designs in order to automate the decision making process. A simplified model of the SLC is analyzed with the fault tree analysis technique in order to compare this results with those obtained with the FCM and show consistency with the results, in order to see that both techniques show similar results even if the approaches are different. © 2011 Published by Elsevier B.V.

Nunez-Carrera A.,Comision Nacional de Seguridad Nuclear y Salvaguardias | Prieto-Guerrero A.,Metropolitan Autonomous University | Espinosa-Paredes G.,Metropolitan Autonomous University
Annals of Nuclear Energy | Year: 2014

This work is a compilation of the current status of the steam dryer performance after the implementation of power uprates in Boiling Water Reactors (BWR). Some Nuclear Power Plants (NPPs) have reported failures and cracking in the steam dryer by acoustic resonances that cause excessive vibration due to the increase of steam flow. The replacement of the steam dryer, structural reinforcement and the connection of Acoustic Side Branches (ASB) are the main solutions adopted in order to avoid mechanical failures. The signal analysis of the vibration of the main steam lines in a typical BWR5, was performed using the Short-Time Fourier Transform (STFT). Signals were collected by the strain gauges located around the main steam lines (MSL). Two scenarios are analyzed; the first, is the signal obtained before the installation of the ASB, and the second, the signal obtained after installation. The results show the effectiveness of the ASB to damp the strong resonances when the steam flow increases, which represents an important reduction in the acoustic loads to the steam dryer. © 2014 Elsevier Ltd. All rights reserved.

Espinosa-Paredes G.,Metropolitan Autonomous University | Verma S.P.,National Autonomous University of Mexico | Vazquez-Rodriguez A.,Metropolitan Autonomous University | Nunez-Carrera A.,Comision Nacional de Seguridad Nuclear y Salvaguardias
Nuclear Engineering and Design | Year: 2010

Our aim was to evaluate the sensitivity and uncertainty of mass flow rate in the core on the performance of natural circulation boiling water reactor (NCBWR). This analysis was carried out through Monte Carlo simulations of sizes up to 40,000, and the size, i.e., repetition of 25,000 was considered as valid for routine applications. A simplified boiling water reactor (SBWR) was used as an application example of Monte Carlo method. The numerical code to simulate the SBWR performance considers a one-dimensional thermo-hydraulics model along with non-equilibrium thermodynamics and non-homogeneous flow approximation, one-dimensional fuel rod heat transfer. The neutron processes were simulated with a point reactor kinetics model with six groups of delayed neutrons. The sensitivity was evaluated in terms of 99% confidence intervals of the mean to understand the range of mean values that may represent the entire statistical population of performance variables. The regression analysis with mass flow rate as the predictor variable showed statistically valid linear correlations for both neutron flux and fuel temperature and quadratic relationship for the void fraction. No statistically valid correlation was observed for the total heat flux as a function of the mass flow rate although heat flux at individual nodes was positively correlated with this variable. These correlations are useful for the study, analysis and design of any NCBWR. The uncertainties were propagated as follows: for 10% change in the mass flow rate in the core, the responses for neutron power, total heat flux, average fuel temperature and average void fraction changed by 8.74%, 7.77%, 2.74% and 0.58%, respectively. © 2010 Elsevier B.V. All rights reserved.

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