http://chineseinput.net/에서 pinyin(병음)방식으로 중국어를 변환할 수 있습니다.
변환된 중국어를 복사하여 사용하시면 됩니다.
- 中文 을 입력하시려면 zhongwen을 입력하시고 space를누르시면됩니다.
- 北京 을 입력하시려면 beijing을 입력하시고 space를 누르시면 됩니다.
RISS 인기검색어
- 검색키워드
- 저자 : FEGHHI, S. A. (검색결과 9 건)
- 무료
- 기관 내 무료
- 유료
-
-
Nazemi, E.,Feghhi, S.A.H.,Roshani, G.H.,Gholipour Peyvandi, R.,Setayeshi, S. Korean Nuclear Society 2016 Nuclear Engineering and Technology Vol.48 No.1
Void fraction is an important parameter in the oil industry. This quantity is necessary for volume rate measurement in multiphase flows. In this study, the void fraction percentage was estimated precisely, independent of the flow regime in gas-liquid two-phase flows by using ${\gamma}-ray$ attenuation and a multilayer perceptron neural network. In all previous studies that implemented a multibeam ${\gamma}-ray$ attenuation technique to determine void fraction independent of the flow regime in two-phase flows, three or more detectors were used while in this study just two NaI detectors were used. Using fewer detectors is of advantage in industrial nuclear gauges because of reduced expense and improved simplicity. In this work, an artificial neural network is also implemented to predict the void fraction percentage independent of the flow regime. To do this, a multilayer perceptron neural network is used for developing the artificial neural network model in MATLAB. The required data for training and testing the network in three different regimes (annular, stratified, and bubbly) were obtained using an experimental setup. Using the technique developed in this work, void fraction percentages were predicted with mean relative error of <1.4%.
-
Ebrahimkhani, Marziye,Hassanzadeh, Mostafa,Feghhi, Sayed Amier Hossian,Masti, Darush Korean Nuclear Society 2016 Nuclear Engineering and Technology Vol.48 No.1
Calculation of the core neutronic parameters is one of the key components in all nuclear reactors. In this research, the energy spectrum and spatial distribution of the neutron flux in a uranium target have been calculated. In addition, sensitivity of the core neutronic parameters in accelerator-driven subcritical advanced liquid metal reactors, such as electron beam energy ($E_e$) and source multiplication coefficient ($k_s$), has been investigated. A Monte Carlo code (MCNPX_2.6) has been used to calculate neutronic parameters such as effective multiplication coefficient ($k_{eff}$), net neutron multiplication (M), neutron yield ($Y_{n/e}$), energy constant gain ($G_0$), energy gain (G), importance of neutron source (${\varphi}^*$), axial and radial distributions of neutron flux, and power peaking factor ($P_{max}/P_{ave}$) in two axial and radial directions of the reactor core for four fuel loading patterns. According to the results, safety margin and accelerator current ($I_e$) have been decreased in the highest case of $k_s$, but G and ${\varphi}^*$ have increased by 88.9% and 21.6%, respectively. In addition, for LP1 loading pattern, with increasing $E_e$ from 100 MeV up to 1 GeV, $Y_{n/e}$ and G improved by 91.09% and 10.21%, and $I_e$ and $P_{acc}$ decreased by 91.05% and 10.57%, respectively. The results indicate that placement of the Np-Pu assemblies on the periphery allows for a consistent $k_{eff}$ because the Np-Pu assemblies experience less burn-up.
-
Parisa Pasangi,Mahmoud Atashbar,Mahmood Mohassel Feghhi 한국전자통신연구원 2019 ETRI Journal Vol.41 No.4
In time division duplex (TDD)‐based multiuser massive multiple input multiple output (MIMO) systems, the uplink channel is estimated and the results are used in downlink for signal detection. Owing to noisy uplink channel estimation, the downlink channel should also be estimated for accurate signal detection. Therefore, recently, a blind method was developed, which assumes the use of a linear high‐power amplifier (HPA) in the base station (BS). In this study, we extend this method to a scenario with a nonlinear HPA in the BS, where the Bussgang decomposition is used for HPA modeling. In the proposed method, the average power of the received signal for each user is a function of channel gain, large‐scale fading, and nonlinear distortion variance. Therefore, the channel gain is estimated, which is required for signal detection. The performance of the proposed method is analyzed theoretically. The simulation results show superior performance of the proposed method compared to that of the other methods in the literature.
-
Marziye Ebrahimkhani,Mostafa Hassanzadeh,Sayed Amier Hossian Feghhi,Darush Masti 한국원자력학회 2016 Nuclear Engineering and Technology Vol.48 No.1
Calculation of the core neutronic parameters is one of the key components in all nuclearreactors. In this research, the energy spectrum and spatial distribution of the neutron fluxin a uranium target have been calculated. In addition, sensitivity of the core neutronicparameters in accelerator-driven subcritical advanced liquid metal reactors, such aselectron beam energy (Ee) and source multiplication coefficient (ks), has been investigated. A Monte Carlo code (MCNPX_2.6) has been used to calculate neutronic parameters such aseffective multiplication coefficient (keff), net neutron multiplication (M), neutron yield (Yn/e), energy constant gain (G0), energy gain (G), importance of neutron source (4*), axial andradial distributions of neutron flux, and power peaking factor (Pmax/Pave) in two axial andradial directions of the reactor core for four fuel loading patterns. According to the results,safety margin and accelerator current (Ie) have been decreased in the highest case of ks, butG and 4* have increased by 88.9% and 21.6%, respectively. In addition, for LP1 loadingpattern, with increasing Ee from 100 MeV up to 1 GeV, Yn/e and G improved by 91.09% and10.21%, and Ie and Pacc decreased by 91.05% and 10.57%, respectively. The results indicatethat placement of the NpePu assemblies on the periphery allows for a consistent keffbecause the NpePu assemblies experience less burn-up.
-
E. Nazemi,S.A.H. FEGHHI,G.H. Roshani,R. Gholipour Peyvandi,S. Setayeshi 한국원자력학회 2016 Nuclear Engineering and Technology Vol.48 No.1
Void fraction is an important parameter in the oil industry. This quantity is necessary forvolume rate measurement in multiphase flows. In this study, the void fraction percentagewas estimated precisely, independent of the flow regime in gaseliquid two-phase flows byusing g-ray attenuation and a multilayer perceptron neural network. In all previous studiesthat implemented a multibeam g-ray attenuation technique to determine void fractionindependent of the flow regime in two-phase flows, three or more detectors were usedwhile in this study just two NaI detectors were used. Using fewer detectors is of advantagein industrial nuclear gauges because of reduced expense and improved simplicity. In thiswork, an artificial neural network is also implemented to predict the void fraction percentageindependent of the flow regime. To do this, a multilayer perceptron neuralnetwork is used for developing the artificial neural network model in MATLAB. Therequired data for training and testing the network in three different regimes (annular,stratified, and bubbly) were obtained using an experimental setup. Using the techniquedeveloped in this work, void fraction percentages were predicted with mean relative errorof <1.4%.
-
S. Jafarikia,S.A.H. FEGHHI 한국원자력학회 2019 Nuclear Engineering and Technology Vol.51 No.1
The purpose of this paper is to study the source term behavior after severe accidents by using a semikineticmodel for simulation and calculation of in-containment activity. The reactor containment specificationand the safety features of the containment under different accident conditions play a great rolein evaluating the in-containment activity. Assuming in-vessel and instantaneous release of radioactivity into the containment, the behavior of incontainmentisotopic activity is studied for noble gasses (Kr and Xe) and the more volatile elements ofiodine, cesium, and aerosols such as Te, Rb and Sr as illustrative examples of source term release underLOCA conditions. The results of the activity removal mechanisms indicates that the impact of volumetricleakage rate for noble gasses is important during the accident, while the influence of deposition on thecontainment surfaces for cesium, mainly iodine isotopes and aerosol has the largest contribution inremoval of activity during evolution of the accident.
-
Z. GHOLAMZADEH,S.A.H. FEGHHI,S.M. MIRVAKILI,A. JOZE-VAZIRI,M. ALIZADEH 한국원자력학회 2015 Nuclear Engineering and Technology Vol.47 No.7
The use of subcritical aqueous homogenous reactors driven by accelerators presents anattractive alternative for producing 99Mo. In this method, the medical isotope productionsystem itself is used to extract 99Mo or other radioisotopes so that there is no need toirradiate common targets. In addition, it can operate at much lower power compared to atraditional reactor to produce the same amount of 99Mo by irradiating targets. In this study,the neutronic performance and 99Mo, 89Sr, and 131I production capacity of a subcriticalaqueous homogenous reactor fueled with low-enriched uranyl nitrate was evaluated usingthe MCNPX code. A proton accelerator with a maximum 30-MeV accelerating power wasused to run the subcritical core. The computational results indicate a good potential for themodeled system to produce the radioisotopes under completely safe conditions because ofthe high negative reactivity coefficients of the modeled core. The results show thatapplication of an optimized beam window material can increase the fission power of theaqueous nitrate fuel up to 80%. This accelerator-based procedure using low enricheduranium nitrate fuel to produce radioisotopes presents a potentially competitive alternativein comparison with the reactor-based or other accelerator-based methods. This systemproduces ~1,500 Ci/wk (~325 6-day Ci) of 99Mo at the end of a cycle.
-
Amini, Moharram,Zamzamian, Seyed Mehrdad,Fadaei, Amir Hossein,Gharib, Morteza,Feghhi, Seyed Amir Hosein Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.10
Applying the available neutron flux for medical and industrial purposes is the most important application of research reactors. The neutron radiography system is used for non-destructive testing (NDT) of materials so that it is one of the main applications of nuclear research reactors. One of these research reactors is the 5 MW pool-type light water research reactor of Tehran (TRR). This work aims to investigate on materials and location of the beam tube (BT) of the TRR radiography system to improve the index parameters of BT. Our results showed that a through-type BT with 20 cm thick carbon neutron filter, 1.2 cm and 9.4 cm of the diameter of inlet (D<sub>1</sub>) and output (D<sub>2</sub>) BT, respectively gives thermal neutron flux almost 25.7, 5.6 and 1.1 times greater than the former design of the TRR (with D<sub>1</sub> = 1.8 cm and D<sub>1</sub> = 9.4 cm), previous design of the TRR with D<sub>1</sub> = 3 cm and D<sub>1</sub> = 9.4 cm, and another design with D<sub>1</sub> = 5 cm and D<sub>1</sub> = 9.4 cm, respectively. Therefore, the design proposed in this paper could be a better alternative to the current BT of the TRR.
연관 검색어 추천
이 검색어로 많이 본 자료
활용도 높은 자료
