Evaluating Wind Energy Potential in Gorgan–Iran Using Two Methods of Weibull Distribution Function

In this study, wind energy characteristics of the, a city in northeast of Iran, measured at 10m height in 2014. The Gorgan airport one hour recorded data extrapolated to 50m height. The data have been statistically analyzed hourly, daily, monthly, seasonally and annually to determine the wind power potential. Weibull distribution function has been used to determine the wind power density and then the potential energy. Standard deviation method and power density method are the methods used to calculate the scaling and shaping parameters of the Weibull distribution function. The annual mean wind power calculated by the standard deviation method and the power density method is 38.98w/m2 and 41.32w/m2, respectively. By comparing the results concluded that the power density method is a better method than the standard deviation method. In addition, Gorgan wind energy potentiality categorized into class 1. So is unsuitable to utilize large wind energy turbine. Article History: Received November 21, 2015; Received in revised form January 15, 2016; Accepted February 10, 2016; Available online How to Cite This Article: Babayani, D., Khaleghi, M., Tashakor, S., and Hashemi-Tilehnoee.,M. (2016) Evaluating wind energy potential in Gorgan–Iran using two methods of Weibull distribution function. Int. Journal of Renewable Energy Development, 5(1), 43-48. http://dx.doi.org/10.14710/ijred.5.1.43-4

Thermal Hydraulic Modeling of Once-Through Steam Generator by Two-Fluid U-Tube Steam Generator Code

The THERMIT U-tube steam generator (THERMIT-UTSG) code was used for evaluation for the parametric study of a scaled once-through pressurized water reactor steam generator (OTSG) made by Babcock & Wilcox. The results of the code were compared to the experimental data of the 19-tube OTSG and a simple heat transfer code that was developed by Osakabe. The main calculated thermodynamic parameters were primary-secondary fluid temperatures, tube wall internal and external temperatures that were subjected to primary and the secondary fluid, and the secondary fluid vapor quality. The assessed code can be used for modeling the OTSGs with some modification. The results of THERMIT-UTSG were in agreement with the experimental results and the prediction of Osakabe’s numerical model

Thermal Hydraulic Modeling of Once-Through Steam Generator by Two-Fluid U-Tube Steam Generator Code

The THERMIT U-tube steam generator (THERMIT-UTSG) code was used for evaluation for the parametric study of a scaled once-through pressurized water reactor steam generator (OTSG) made by Babcock & Wilcox. The results of the code were compared to the experimental data of the 19-tube OTSG and a simple heat transfer code that was developed by Osakabe. The main calculated thermodynamic parameters were primary-secondary fluid temperatures, tube wall internal and external temperatures that were subjected to primary and the secondary fluid, and the secondary fluid vapor quality. The assessed code can be used for modeling the OTSGs with some modification. The results of THERMIT-UTSG were in agreement with the experimental results and the prediction of Osakabe's numerical model

Effect of inclined magnetic field on the entropy generation in an annulus filled with NEPCM suspension

The encapsulation technique of phase change materials in the nanodimension is an innovative approach to improve the heat transfer capability and solve the issues of corrosion during the melting process. (is new type of nanoparticle is suspended in base fluids call NEPCMs, nanoencapsulated phase change materials. (e goal of this work is to analyze the impacts of pertinent parameters on the free convection and entropy generation in an elliptical-shaped enclosure filled with NEPCMs by considering the effect of an inclined magnetic field. To reach the goal, the governing equations (energy, momentum, and mass conservation) are solved numerically by CVFEM. Currently, to overcome the low heat transfer problem of phase change material, the NEPCM suspension is used for industrial applications. Validation of results shows that they are acceptable.(eresults reveal that the values of Nuave descend with ascending Ha while Ngen has a maximum at Ha 16. Also, the value of NT,MF increases with ascending Ha. (e values of Nuave and Ngen depend on nondimensional fusion temperature where good performance is seen in the range of 0.35 < θf < 0.6. Also, Nuave increases 19.9% and ECOP increases 28.8% whereasNgen descends 6.9% when ϕ ascends from 0 to 0.06 at θf 0.5. Nuave decreases 4.95% while Ngen increases by 8.65% when Ste increases from 0.2 to 0.7 at θf 0.35.http://www.hindawi.com/journals/mpeam2022Mechanical and Aeronautical Engineerin

Impact of fusion temperature on hydrothermal features of flow within an annulus loaded with nanoencapsulated phase change materials (NEPCMs) during natural convection process

Please read abstract in the article.http://www.hindawi.com/journals/mpe/am2022Mechanical and Aeronautical Engineerin

Sensitivity analysis of thermal-hydraulic parameters to study the corrosion intensity in nuclear power plant steam generators

The failure of steam generators (SGs) due to corrosion is one of the most important problems in power plants. Impurities usually accumulate in the hot sides of SG and form deposits on the SG surfaces. In this paper, the sensitivity analysis of the accumulation of water impurities in the heat exchangers of nuclear power plants is presented. The convection-diffusion equation of the liquid phase on the heated surfaces is derived and then solved by the finite volume method. Also, the effects of the thermal–hydraulic parameters in the form of dimensionless numbers, such as Peq, Peu, kp(relative solubility of impurity between the steam and water) on the impurities concentration are studied. Keywords: Corrosion, Steam generators, Impurities, Hideout and retur

Thermal-hydraulic analysis of VVER-1000 residual heat removal system using RELAP5 code, an evaluation at the boundary of reactor repair mode

Removing the residual heat from a nuclear reactor is an important safety aspect of thermal hydraulic analysis. In this study, a typical VVER-1000 reactor residual heat removal system has been evaluated using RELAP5 thermal hydraulic loop code during cool-down. Reactor cooling down starts from hot state temperature and then continues to the cool-down stages with 130 °C and 70 °C, respectively. The second stage of cooling down is the boundary of the reactor repair condition. Main cooling pump head, steam generator (SG) water level, system pressure, the level of coolant in the pressurizer (PRZ), and the temperature of fuel element are examined in a steady state condition. PRZ level, primary circuit and secondary circuit pressure/temperature, and SG water level are evaluated during 32,000 s after cool-down scenario. By comparison, it concluded that the results of RELAP5 code are in agreement with plant experimental data and final safety analysis report (FSAR). Thus, it is proved that the studied reactor is capable to remove the residual heat generated during shutdown. Moreover, RELAP5 is properly recommended for analysis of the VVER-1000 pressurized water reactor during cool-down. Keywords: RELAP5, VVER-1000, BNPP, FSAR, Residual heat removal, RH