Wyniki wyszukiwania - BazTech

1

Design and simulation of a multienergy gamma ray absorptiometry system for multiphase flow metering with accurate void fraction and water-liquid ratio approximation

Khayat Omid, Afarideh Hossein

Nukleonika

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2019

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Vol. 64, No. 1

19--29

EN

Multiphase flow meters are used to measure the water-liquid ratio (WLR) and void fraction in a multiphase fluid stream pipeline. In the present study, a system of multiphase flow measurement has been designed by application of three thallium-doped sodium iodide scintillators and a radioactive source of 133Ba simulated by Monte Carlo N-particle (MCNP) transport code. In order to capture radiations passing across the pipe, two direct detectors have been installed on opposite sides of the radioactive source. Another detector has been placed perpendicular to the transmission beam emitted from the 133Ba source to receive radiations scattered from the fluid flow. Simulation was done by the MCNP code for different volumetric fractions of water, oil, and gas phases for two types of flow regimes, namely, homogeneous and annular; training and validation data have been provided for the artificial neural network (ANN) to develop a computation model for pattern recognition. Depending on applications of the neural system, several structures of ANNs are used in the current paper to model the flow measurement relations, while the detector outputs are considered as the input parameters of the neural networks. The first, second, and third structures benefit from two, three, and five multilayer perceptron neural networks, respectively. Increasing the number of ANNs makes the system more complicated and decreases the available data; however, it increases the accuracy of estimation of WLR and gas void fraction. According to the results, the maximum relative difference was observed in the scattering detector. It was clear that transmission detectors would demonstrate the difference between the flow regimes as well. It is necessary to note that the error calculated by the MCNP simulator is <0.5% for the direct detectors (TR1 and TR2). Due to the difference between the data of the two flow regimes and the errors of data in the simulation codes of the MCNP, it was possible to separate these flow regimes. The effect of changing WLR on the efficiency for a constant void fraction confi rms a considerable variance in the results of annular and homogeneous flow s occurring in the scattering detector. There is a similar trend for the void fraction; hence, one can easily distinguish changes in efficiency due to the WLR. Analysis of the simulation results revealed that in the proposed structure of the multiphase flow meter and the computation model used for simulation, the two flow regimes are simply distinguishable.

2

A New Approach in Coal Mine Exploration Using Cosmic Ray Muons

Darijani R., Negarestani A., Rezaie M. R., Fatemi S. J., Akhond A.

Acta Geophysica

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2016

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Vol. 64, no. 4

1034--1050

EN

Muon radiography is a technique that uses cosmic ray muons to image the interior of large scale geological structures. The muon absorption in matter is the most important parameter in cosmic ray muon radiography. Cosmic ray muon radiography is similar to X-ray radiography. The main aim in this survey is the simulation of the muon radiography for exploration of mines. So, the production source, tracking, and detection of cosmic ray muons were simulated by MCNPX code. For this purpose, the input data of the source card in MCNPX code were extracted from the muon energy spectrum at sea level. In addition, the other input data such as average density and thickness of layers that were used in this code are the measured data from Pabdana (Kerman, Iran) coal mines. The average thickness and density of these layers in the coal mines are from 2 to 4 m and 1.3 gr/cm3 , respectively. To increase the spatial resolution, a detector was placed inside the mountain. The results indicated that using this approach, the layers with minimum thickness about 2.5 m can be identified.

3

MCNP calculations of neutron emission anisotropy caused by the GIT-12 hardware

Šíla O., Klír D., Řezáč K., Cikhardtova B., Cikhardt J.

Nukleonika

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2015

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Vol. 60, No. 2

323--326

EN

The MCNP6 and MCNPX calculations for the GIT-12 device in Tomsk were performed to determine the influence of the gas-puff hardware on the neutron emission anisotropy and the neutron scattering rate. A monoenergetic 2.45 MeV neutron source and F1 and F6 tallies were declared in the simulation input. A comparison between MCNP results and the measured data was made. Differences between MCNPX and MCNP6 output data were investigated. In the experiment, two nTOF scintillation detectors with the Bicron BC-408 scintillator were used to measure the neutron waveform. Four bubble BD-PND detectors were used to estimate the amount of neutrons in different places around the neutron source.

4

Comparison of simple design of sodium and lead cooled fast reactor cores

Mazgaj P., Darnowski P., Gurgacz S., Lipka M., Dziubanii K.

Journal of Power Technologies

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2014

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Vol. 94, nr 5

16--25

EN

This report presents the results of a numerical simulation of thermal hydraulics processes in a liquid metal cooled fast reactor core, combined with simple neutron population computing for an infinite pin cell lattice. Two types of coolant were studied: liquid sodium and liquid lead, with all requirements regarding safety conditions observed. Temperature distributions along the cooling channel and distributions in the radial direction were prepared, then criticality calculations were performed for MOX fuel using MCNP Monte Carlo code.

5

Determination of the geometry function for a brachytherapy seed, comparing MCNP results with TG-43U1 analytical approximations

Raisali G., Ghonchehnazi M. G., Shokrani P., Sadeghi M.

Nukleonika

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2008

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Vol. 53, No. 2

45-49

EN

Geometry function is the only dosimetry parameter of a brachytherapy source seed, introduced in TG-43U1 protocol which is determined using calculational methods rather than physical measurement. In order to evaluate the accuracy of point and line source approximations, for calculation of the geometry function, the MCNP computer code has been used for a typical brachytherapy seed and the results have been compared. The MCNP has been used to simulate the geometry and activity distribution of a Pd-103 seed in order to calculate the geometry function for various angles and distances from the source. The comparison of results shows that at distances close to the source, the values predicted with different methods are not in agreement. The difference between the MCNP calculations and line approximation for small angles from ? = 0 to 15° is about 27% at 0.25 cm from the seed center. This difference is so much higher for point source approximation (up to a factor of 3) even up to distances of 0.5 cm from the source. As ? increases, the difference between MCNP and approximate methods is reduced. Therefore, for small distances from brachytherapy seeds, it is recommended to calculate the geometry function using more detailed methods instead of point and linear source approximations. This will provide more accurate results for other TG-43U1 dosimetry parameters such as radial dose function or anisotropy function which for some points are calculated via interpolation or extrapolation of the available discrete dosimetry data.

6

Numerical optimisation of the fission-converter and the filter/moderator arrangement for the Boron Neutron Capture Therapy

Tracz G., Dąbkowski L., Dworak D., Pytel K., Woźnicka U.

Nukleonika

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2003

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Vol. 48, nr 4

177-185

EN

The paper presents results of the numerical modelling of the fission-converter-based epithermal neutron source designed for a BNCT (Boron Neutron Capture Therapy) facility to be located at the Polish research nuclear reactor MARIA at Świerk. The unique design of the fission converter has been proposed due to a specific geometrical surrounding of the reactor. The filter/moderator arrangement has been optimised in order to moderate fission neutrons to epithermal energies and to get rid of both fast neutrons and photons from the therapeutic beam. The selected filter/moderator set-up ensures both the high epithermal neutron flux and the suitably low level of beam contamination. The elimination of photons originated in the reactor core is an exceptional advantage of the proposed design. It brings one order of magnitude lower gamma radiation dose than the permissible dose in such a type of therapeutic facility is required. The MCNP and FLUKA codes have been used for the computations.