Neutron unfolding using MCNPX code and SAND-II algorithm
یاسر
کاسه ساز
پژوهشگر
author
Ablofazl
Heydarzadeh
PNU
author
Saeed
Mohammadi
PNU
author
text
article
2017
per
Neutron spectroscopy is very important in development of neutron applications. The most commonly used method to measure the neutron energy spectrum is the threshold foil activation method and using an unfolding code such as SAND-II code. The main limitation of this code is that the geometry of the source and measurement setup could not define in the code. In this study, to eliminate this limitation, a new unfolding code has been developed by iterative algorithm in SAND-II and MCNPX code. The full geometry of the measurement setup including the source and foil can be simulated by MCNPX code. In the proposed code, the modified iteration algorithm used in SAND-II is used. The results of the research show that the arrangement of foils in front of the neutron source is effective in the amount of saturation activities of them. Also, using the information obtained by our proposed code, the spectrum unfolded by this code has a good agreement with other neutron spectrum unfolding methods.
Journal of Radiation and Nuclear Technology
University of Guilan
2423-6616
4
v.
3
no.
2017
1
9
https://jrnt.guilan.ac.ir/article_2820_d41d8cd98f00b204e9800998ecf8427e.pdf
dx.doi.org/10.22124/jrnt.2018.9511.1075
Aboard Dose Assessment in Dual Energy X-ray Vehicle Inspection by Use of Experimental Measurement in Human Rando Phantom
Seyed Mohammad
Hasheminejad
Physics Dept., Islamic Azad University, Tehran, Iran
author
Mohammad Mahdi
Mojarad Kahani
Radiation Medicine Engineering Dept., Shahid Beheshti University, Tehran, Iran
author
Hossein
Jafari
Physics and Nuclear Engineering Dept., Amirkabir University of Technology, Tehran, Iran
author
Hamid
Shafaei douk
Radiation Medicine Engineering Dept., Shahid Beheshti University, Tehran, Iran
author
Morteza
Azarbadegan
Physics Dept., Islamic Azad University, Tehran, Iran
author
Samaneh
Hashemi
radiation medicine department, Shahid Beheshti university, Tehran, Iran
author
text
article
2017
per
One of the most effective and comprehensive ways to deal with security risks as well as smuggling of contraband including drugs is using Dual Energy X-ray inspection devices with the capability of materials classification. An important parameter in relation to the use of these devices in the vehicle inspection is the car driver and sensitive body organs dose which can lead to person’s diseases caused by radiation. The aim of this study is the risk assessment for car driver in each scan and also determination of maximum number passes through inspection device for a specified time based on individual doses per scan and the ANSI standard. By using a homespun and programmed car inspection dual-energy X-ray as well as using a Rando phantom of a man and TLD dosimeters, dosimetry was performed to assess the doses of overall and sensitive body organs at three energy situations of the X-ray tube, 155, 175 and 195 kVp. The results show that maximum dose is absorbed in the stomach and the mean total dose of car driver for 155, 175 and 195 kVp energy is 0.276. 0.315 and 0.338μSv respectively. With respect to the overall mean car driver dose (the 0.31μSv) and according to the ANSI standard, machine is known in the category of devices with limited utility. Anyone can pass through this machine in maximum for a year, a month and a week 806, 66 and 15 times respectively and more scan could lead to radiation biohazards according to ANSI standard.
Journal of Radiation and Nuclear Technology
University of Guilan
2423-6616
4
v.
3
no.
2017
10
18
https://jrnt.guilan.ac.ir/article_2821_d41d8cd98f00b204e9800998ecf8427e.pdf
dx.doi.org/10.22124/jrnt.2018.8419.1063
Design, Fabrication and Assessment of proportional counter in current and sealed gas mode
jamshid
Soltani-Nabipour
Department of medical Radiation Engineering, Faculty of computer and Technology, Islamic Azad University, Parand Branch, Tehran-Iran
author
farideh
Sadeghi
Departement of medical Radiation Engineering, Faculty of computer and Information Technology, Islamic Azad University, Parand Branch,Tehran-Iran
author
text
article
2017
per
The proportional counter is a type of gaseous ionization detector device used to measure particles of ionizing radiation. The key feature is its ability to measure the energy of incident radiation, by producing a detector output that is proportional to the radiation energy. It is widely used where energy levels of incident radiation must be known, such as in the discrimination between alpha and beta particles, or accurate measurement of X-ray radiation dose. In this study some sample of proportional counters in mode of current gas and sealed gas were designed and constructed. The advantages and disadvantages of each one were studied. In this study, the plateau curve and functional of these detectors were studied and measured by Radioisotopes sources 133Ba, 241 Am, 152. The results of data shows that the energy discrimination for photon radiation in these detectors are good and it can be concluded that these kind of detectors can be constructed in our country with better results.
Journal of Radiation and Nuclear Technology
University of Guilan
2423-6616
4
v.
3
no.
2017
19
26
https://jrnt.guilan.ac.ir/article_2822_d41d8cd98f00b204e9800998ecf8427e.pdf
dx.doi.org/10.22124/jrnt.2018.9191.1069
X-ray tube design based on current and voltage of X-ray tube
seyed morteza
esmaeili
Application Group Beams-Faculty of Nuclear Engineering-Shahid Beheshti University of Tehran
author
Ruhollah
Ghaderi
Department of Application of Beams, Faculty of Nuclear Engineering, Shahid Beheshti University, Tehran, Iran
author
text
article
2017
per
Cargo inspection systems using X-Ray that is widely used in airports and transportation systems are play an important role in supplying the countries security. In this study, the X-Ray production system and its body has been simulated using MCNPX2.6 Monte Carlo code. In X-Ray lamp of this system, the energy of electrons emitted to anode is 160 keV and its current is 0.6 mA. The tungsten anode is located in glass chamber of X-ray lamp with distance of 5.3 from cathode. The considered tunnel’s dimensions are 120cm in Z (object path), 100cm in Y (align with X-ray lamp) and 100cm in X. the body of tunnel has 1mm lead and 2mm iron. The dose rate in 1cm in and out of the tunnel (leaked dose) is calculated using MCNPX2.6. The most amount of dose is in center of the tunnel exactly below the X-ray lamp and dose in 4 sides of outer of the tunnel is less than 5 μSv/h. these results show that the 1mm thickness of lead and 2mm thickness of iron is appropriate for tunnel body
Journal of Radiation and Nuclear Technology
University of Guilan
2423-6616
4
v.
3
no.
2017
27
33
https://jrnt.guilan.ac.ir/article_2825_d41d8cd98f00b204e9800998ecf8427e.pdf
dx.doi.org/10.22124/jrnt.2018.9396.1072
Study of the role of dopants and annealing temperature in LiF:Mg,Cu,P
akram
yahyaabadi
دانشجو
author
Falamarz
Torkzadeh
Atom Energy Organization of Iran, Nuclear Science and Technology Research Institute, Avenue North Kargar, Tehran, Iran, PO Box 14395-836
author
dariush
Rezaeyochbelagh
Nuclear Engineering & Physics Department, Amirkabir University of Technology, Tehran, Iran
author
text
article
2017
per
The LiF:Mg,Cu,P thermoluminesence dosimeter (TLD) is a tissue equivalent material with high sensitivity. This dosimeter is widely utilized in thermolumincence (TL) dosimetry. The TL sensitivity of LiF:Mg,Cu,P are affected by changes in the dopant concentrations. In this study, LiF thermoluminescent materials with Mg, Cu and P dopants were prepared in powder form and their sensitivities were examined to different annealing temperatures within 240-400 °C for 10 min. It was found that thermoluminescence intensity decreases with increasing annealing temperature. Also, the optimum concentration of the Cu and the role of the dopants in the LiF:Mg,Cu,Ag material were investigated. TL intensities have an increasing trend with increasing Cu concentration to 0.05 mol% and then reduce. The results of this study indicate that the each of the three dopants Mg, Cu and P appears to play a role in the presence of each other for the decreasing competitor centers and enhancing thermoluminescent sensitivity.
Journal of Radiation and Nuclear Technology
University of Guilan
2423-6616
4
v.
3
no.
2017
34
41
https://jrnt.guilan.ac.ir/article_2824_d41d8cd98f00b204e9800998ecf8427e.pdf
dx.doi.org/10.22124/jrnt.2018.9223.1070
Calculation of neutronic and kinetic parameters in Isfahan Miniature Neutron Source Reactor using Monte Carlo method and comparison with the results SAR
M.
Ghaed
Department of Nuclear Engineering, Islamic Azad University Science & Research Bosher Branch, Bosher-Iran
author
Mostafa
Hasanzadeh
انرژی اتمی ایران
author
S.A.H.
Feghhi
Radiation Application Department, Nuclear Engineering Faculty, Shahid Beheshti University, Tehran-Iran
author
text
article
2017
per
Kinetic and neutronic parameters have an important role in reactors dynamic behavior analysis. Some of these parameters in nuclear reactors are such as effective multiplication factor (keff), reactivity (ρ), neutron flux as well as power spatial distributions, effective delayed neutron fraction (βeff) and prompt neutron lifetime (Lp). In the current work, analysis and calculation of the kinetic and neutronic parameters are performed using MCNPX code, slope fit and perturbation methods in Isfahan Miniature Neutron Source Reactor (MNSR). According to results, relative difference between the results of MCNPX code and the reference values in calculating of the reactivity and effective delayed neutron fraction are about 0.5% and 2.1%, respectively. The relative difference between the results of the slope fit and perturbation methods with the reference values in calculating of the prompt neutron lifetime are about 5.0%, 9.5% and 8.5%, respectively. Therefore, the results of this research show that the MCNPX code is suitable for calculating of the reactor kinetic parameters such as effective delayed neutrons fraction, while the perturbation method is a simple and convenient method for calculating of the prompt neutron lifetime.
Journal of Radiation and Nuclear Technology
University of Guilan
2423-6616
4
v.
3
no.
2017
42
50
https://jrnt.guilan.ac.ir/article_2823_d41d8cd98f00b204e9800998ecf8427e.pdf
dx.doi.org/10.22124/jrnt.2018.6623.1045