Pengaruh Atribut Produk Wisata dan EWOM di Jejaring Sosial Instagram terhadap Keputusan Berkunjung Objek Wisata Saloka

Penelitian ini bertujuan untuk mengetahui pengaruh EWOM di jejaring sosial instagram dan atribut produk wisata terhadap keputusan berkunjung. Dalam penelitian ini ada 2 (dua) hipotesa yang diuji. Peneliti menggunakan uji linear berganda dengan alat uji IBM SPSS Statistic 24 untuk menguji 200 kuesioner yang telah diisi oleh pengujung yang telah berkunjung ke Saloka. Hasil dari penelitian ini menunjukkan bahwa EWOM di jejaring sosial instagram dan atribut produk wisata secara bersama-sama maupun secara terpisah berpengaruh positifsignifikan

Análisis forense del caso DFRWS 2008 Rodeo

Como es de saber, la informática forense es una disciplina en la cual se encarga de proteger la información, con el fin de poder examinar la investigación encontrada en un equipo o dispositivo, y así se podría utilizar como evidencia al momento de estar en un tribunal de justicia; con base a lo dicho anteriormente se desarrolló un análisis de un caso forense llamado DFWS Rodeo en la cual dice que existieron posibles perjuicios de un sistema de información, y la red de la compañía en la cual el acusado es el señor Vogon, en donde el señor Vogon estaba disgustado con la compañía, en donde a medida que se iba investigando el caso supimos que el señor Vogon salió nervioso y disgustado de una entrevista que le hizo el área de recursos humanos, en donde después de lo sucedido se procedió a decomisar un pendrive y una cámara que se encontraba en el escritorio del señor Vogon. Con el fin de poder analizar dichos dispositivos, se utilizó un programa llamado Autopsy cuya función es analizar los dispositivos de almacenamiento, en la cual se analizó un disco duro y un pendrive. Para esto se realizó también un informe donde se encuentra la información más concreta sobre las evidencias encontradas en el caso

Problema de programacion de operaciones y herramientas en un sistema de manufactura flexible: heuristica de carga fase i

Esta fase toma los resultados obtenidos en la fase I de la Heurística de Carga, cuyo objetivo es el de asignar operaciones y herramientas a máquinas dentro de un Sistema de Manufactura Flexible (FMS) con el objetivo de balancear cargas de trabajo entre tipos de máquinas, generar flexibilidad en las rutas de producción y disminuir el intercambio de material entre máquinas. La Fase I se enfoca en la asignación de operaciones a tipos de máquinas de tal manera que se logre un balance de las cargas de trabajo entre estos. Por otro lado la Fase II busca lograr los objetivos de disminución de los intercambios de material y la flexibilización de las rutas de producción

Salmonella, Coliformes totais e fecais em queijo Minas artesanal comercializado em feiras-livre da cidade de Uberlândia-MG

Trabalho de Conclusão de Curso (Graduação)O queijo minas artesanal é produzido com leite cru podendo trazer prejuízos à saúde do consumidor devido a presença de microorganismos patogênicos. Com o objetivo de verificar a qualidade sanitária dos queijos Minas artesanais comercializados em Uberlândia-MG, foram coletadas em feiras-livre, nos meses de julho a outubro de 2004, 23 amostras de queijo. Foram realizadas análises do teor de umidade (%), quantificação de coliformes totais e fecais (NMP.g1) e presença/ausência de Salmonella sp. Os resultados das contagens de coliformes foram correlacionados com os teores de umidade. A umidade média dos queijos foi 41,01%. O numero de amostras contaminadas por coliformes totais e fecais foram de 52,1% (1E23) e 47,8% (11/23), respectivamente, não sendo detectada a presença de Salmonella sp. Das amostras analisadas, 65,2% (15/23) apresentaram resultados acima dos permitidos pela legislação para o grupo conforme, sendo consideradas impróprias para o consumo. O coeficiente de correlação calculado para umidade e contagens de coliformes totais e fecais foi de 0,88 e 0,89, respectivamente (p<0,10). Os resultados obtidos indicam que os queijos Minas artesanais comercializados nas feiras livres de Uberlândia-MG devem merecer atenção dos órgãos de saúde pública e dos consumidores por oferecerem riscos à saúde pública

Uncertainty and Sensitivity Analysis Applied to the Validation of BWR Bundle Thermal-Hydraulic Calculations

In recent years, more realistic safety analyses of nuclear reactors have been based on best estimate (BE) computercodes. The need to validate and refine BE codes that are used in the predictions of relevant reactor safetyparameters, led to the organization of international benchmarks based on high quality experimental data. TheOECD/NRC BWR Full‐Size Fine‐Mesh Bundle Test (BFBT) benchmark offers a good opportunity to assess theaccuracy of thermal‐hydraulic codes in predicting, among other parameters, single and two phase bundle pressuredrops, cross‐sectional averaged void fraction distributions and critical powers under a wide range of systemconditions. The BFBT is based on a multi‐rod assembly integral test facility which is able to simulate the highpressure, high temperature fluid conditions found in BWRs through electrically‐heated rod bundles. Since codeaccuracy is unavoidably affected by models and experimental uncertainties, an uncertainty analysis is fundamentalin order to have a complete validation study.This work has two main objectives. The first one is to enhance the validation process of the thermal‐hydraulicfeatures of the Westinghouse code POLCA‐T. This is achieved by computing a quantitative validation limit based onstatistical uncertainty analysis. This validation theory is applied to some of the benchmark cases of the followingmacroscopic BFBT exercises: 1) Single and two phase bundle pressure drops, 2) Steady‐state cross‐sectionalaveraged void fraction, 3) Transient cross‐sectional averaged void fraction and 4) Steady‐state critical power tests.Sensitivity analysis is also performed to identify the most important uncertain parameters for each exercise.The second objective consists in showing the clear advantages of using the quasi‐random Latin HypercubeSampling (LHS) strategy over simple random sampling (SRS). LHS allows a much better coverage of the inputuncertainties than SRS because it densely stratifies across the range of each input probability distribution. The aimhere is to compare both uncertainty analyses on the BWR assembly void axial profile prediction in steady‐state,and on the transient void fraction prediction at a certain axial level coming from a simulated re‐circulation pumptrip scenario. It is shown that the replicated void fraction mean (either in steady‐state or transient conditions) hasless variability when using LHS than SRS for the same number of calculations (i.e. same input space sample size)even if the resulting void fraction axial profiles are non‐monotonic. It is also shown that the void fractionuncertainty limits achieved with SRS by running 458 calculations (sample size required to cover 95% of 8 uncertaininput parameters with a 95% confidence), result in the same uncertainty limits achieved by LHS with only 100calculations. These are thus clear indications on the advantages of using LHS.Finally, the present study contributes to a realistic analysis of nuclear reactors, in the sense that the uncertaintiesof important BWR parameters at a bundle level are assessed.Keywords: Thermal‐hydraulic codes, uncertainty and sensitivity analysis, BFBT benchmark, Latin Hypercubesampling, simple random sampling, reactor safety analysi

Uncertainty and sensitivity analysis applied to LWR neutronic and thermal-hydraulic calculations

The deterministic modeling of LWRs begins with the computation of energy‐collapsed and homogenizedmacroscopic cross‐sections by means of a lattice code. Once these parameters are functionalized as a functionof the reactor state variables and discretized in space, they are used as input variables by core simulators inorder to calculate the spatial distribution of the neutron flux and thus, the spatial distribution of the power.Once the power is determined, the thermal‐hydraulic variables are updated, and the process repeated untilconvergence. This thesis is divided in three different parts related to the possible neutronic and thermalhydraulicmodeling strategies. In the first part, microscopic cross‐section uncertainties based on two modernnuclear data libraries such as JENDL‐4 and ENDF/B‐VII.1 were derived in multi‐group format. These werepropagated through lattice calculations in order to perform uncertainty analysis on the infinite neutronmultiplication factor (\u100747\u100bb6\u10123b, and on two‐group homogenized macroscopic cross‐sections corresponding to a PWRfuel segment. The aim is to compare the uncertainty assessment on \u100747\u100bb6 and on the macroscopic cross‐sectionswhen the different nuclear libraries are employed. It was found that the computed uncertainties based onJENDL‐4 are much higher than the computed uncertainties based on ENDF/B‐VII.1. A sensitivity analysisshowed that the multi‐group variances of the Uranium‐235 fission reaction based on JENDL‐4 are very high,being this the main reason of the observed large discrepancies in the different uncertainty assessments.In the second part of the thesis, two types of uncertainty analyses were performed on core simulations. Thefirst one corresponds to the forward approach of input uncertainty propagation, where the input uncertainspace formed by the nodal two‐group macroscopic cross sections and diffusion coefficients is sampled bothwith SRS and LHS. The possible ranges of variation of such an input space are based on data from a depletioncalculation corresponding to the cycle 26 of the Swedish Ringhals‐1 BWR. The aim of this study is to comparethe efficiency of the uncertainty assessment performed on the nodal thermal flux when SRS and LHS areemployed. On the other hand, in the second type of uncertainty analysis presented in this chapter,discrepancies between spatial measured and calculated fluxes in Ringhals‐1 are used to perform an inverseuncertainty analysis on the spatial dependence of the different core parameters. This analysis is carried outusing Bayesian statistics, where, for a certain cycle, the frequency distributions of macroscopic cross‐sectionsand diffusion coefficients at every assembly node are updated based on the error distribution of the spatialthermal flux. Emphasis was made on performing uncertainty analysis as well on the coefficients of a nodalcross‐section model. Although a very simple model was derived, the aim is to propose an uncertaintyassessment based on replicated sampling techniques such as the general bootstrap method.Finally, in the third part of the thesis, uncertainty and sensitivity analyses were applied to thermal‐hydrauliccalculations. The objective is to show that when experimental data are available, uncertainty analysis can beused in the validation process of a BE code. Quantitative limits based on a statistical theory were computed tovalidate code thermal‐hydraulic features in predicting pressure drop, void fraction and critical heat flux basedon the macroscopic exercises of the OECD/NRC BWR Full‐Size Fine‐Mesh Bundle Test (BFBT) benchmark.The present study performs a realistic analysis of nuclear reactors, particularly in the uncertainty prediction ofimportant neutronic and thermal‐hydraulic parameters of light water reactors