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PubMed | National University of Colombia, Brazilian Radiological Protection and Dosimetry Institute (IRD), Eletrobras, Hospital San Juan Of Dios and 14 more.
Type: Journal Article | Journal: Radiation protection dosimetry | Year: 2016

Internal dosimetry intercomparisons are essential for the verification of applied models and the consistency of results. To that aim, the First Regional Intercomparison was organised in 2005, and that results led to the Second Regional Intercomparison Exercise in 2013, which was organised in the frame of the RLA 9/066 and coordinated by Autoridad Regulatoria Nuclear of Argentina. Four simulated cases covering intakes of (131)I, (137)Cs and Tritium were proposed. Ninteen centres from thirteen different countries participated in this exercise. This paper analyses the participants results in this second exercise in order to test their skills and acquired knowledge, particularly in the application of the IDEAS Guidelines. It is important to highlight the increased number of countries that participated in this exercise compared with the first one and, furthermore, the improvement in the overall performance. The impact of the International Atomic Energy Agency (IAEA) Projects since 2003 has led to a significant enhancement of internal dosimetry capabilities that strengthen the radiation protection of workers.


Allison C.M.,Innovative Systems Software | Hohorst J.K.,Innovative Systems Software | Allison B.S.,Innovative Systems Software | Konjarek D.,ENCONET | And 4 more authors.
Science and Technology of Nuclear Installations | Year: 2012

Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1-3. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM. The calculations used a detailed RELAP/SCDAPSIM model of the Laguna Verde BWR vessel and related reactor cooling systems. The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy Agency on March 21 to support their emergency response team and later to our Japanese members to support their Fukushima Daiichi specific analysis and model development. Copyright © 2012 C. M. Allison et al.


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.


Romero-Paredes H.,Metropolitan Autonomous University | Vazquez Rodriguez A.,Metropolitan Autonomous University | Espinosa Paredes G.,Metropolitan Autonomous University | Villafan Vidales H.I.,National Autonomous University of Mexico | And 2 more authors.
Applied Thermal Engineering | Year: 2014

In this work the exergy and anergy analysis is applied to the I-S cycle based on a nuclear power for hydrogen production. The exergy losses represent the irreversibility of the process called anergy. The evaluation of anergy is done to both internal anergy (endoanergy) and external anergy (exoanergy) separately for each stage analysed of the S-I process for hydrogen production. The S-I process used as thermal energy source energy from a high temperature nuclear reactor G III and G IV. The separately anergy analysis method is a powerful tool to recognize the internal and external exergy destroyed during a process. With this analysis we found that the overall exergy efficiency of the cycle is rated 37.64%, where the process that generated more anergy is the distillation of aqueous sulphuric acid and the part of the system that generates greater anergy is the exchange of heat between the nuclear fuel rods and helium used in the different processes that require thermal energy. © 2014 Elsevier Ltd.


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.


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.


Polo-Labarrios M.A.,Comision Nacional de Seguridad Nuclear y Salvaguardias | Espinosa-Martinez E.-G.,Metropolitan Autonomous University | Quezada-Garcia S.,Metropolitan Autonomous University | Varela-Ham J.R.,Metropolitan Autonomous University | Espinosa-Paredes G.,Metropolitan Autonomous University
Annals of Nuclear Energy | Year: 2014

The Classical Neutron Point-Kinetic (CNPK) equations are a system of stiff nonlinear ordinary differential equations for the neutron density, which have been subject of countless studies and applications with different approaches in the last seventy years. In this paper we carry out the numerical analysis of the Fractional Neutron Point-Kinetics (FNPK) model for two simple cases of reactivity insertion processes: Case (1) Ramp insertion of reactivity, and Case (2) Sinusoidal form. The FNPK model considers a relaxation time associated with a rapid variation in the neutron flux density, which is considered in the differential operator of fractional order, it is known as anomalous diffusion exponent. Different values of the relaxation time with one-group of delayed neutron precursors were used for this numerical analysis. The results of the neutron flux density with the FNPK model were compared with the CNPK equations for ramp and sinusoidal reactivity insertion processes. In both cases, the neutron density behavior described by the first model, for different relaxation times and anomalous diffusion coefficient values, over-predicts the behavior obtained with the CNPK equations. © 2014 Elsevier Ltd. All rights reserved.


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

The loss-of-coolant accident (LOCA) simulation in the boiling water reactor (BWR) of Laguna Verde Nuclear Power Plant (LVNPP) at 105% of rated power is analyzed in this work. The LVNPP model was developed using RELAP/SCDAPSIM code. The lack of cooling water after the LOCA gets to the LVNPP to melting of the core that exceeds the design basis of the nuclear power plant (NPP) sufficiently to cause failure of structures, materials, and systems that are needed to ensure proper cooling of the reactor core by normal means. Faced with a severe accident, the first response is to maintain the reactor core cooling by any means available, but in order to carry out such an attempt is necessary to understand fully the progression of core damage, since such action has effects that may be decisive in accident progression. The simulation considers a LOCA in the recirculation loop of the reactor with and without cooling water injection. During the progression of core damage, we analyze the cooling water injection at different times and the results show that there are significant differences in the level of core damage and hydrogen production, among other variables analyzed such as maximum surface temperature, fission products released, and debris bed height. Copyright © 2012 Gilberto Espinosa-Paredes et al.


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.

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