• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4231-4235
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• 2022

Considering the importance of deuterium in nuclear science including medical and industrial researches such as (BNCT) and nuclear reactors respectively, it is important to study various possible ways in addition to common methods for measuring its concentration. This study is an effort to measure deuterium concentration using PGNAA. The main idea is to calculate the area under 2.23 MeV gamma-rays photo peak resulting from neutron collision with Hydrogen atoms which are in mix with deuterium in samples. The study carried out by both simulation and experiment. Monte Carlo MCNPX2.6 code has been used for simulation and based on its acceptable results an experimental setup has been arranged. The coordination of results was in the range of R = 0.99 and R = 0.98 in simulation and experiment respectively. The accuracy of the study has been investigated by measuring the concentration of an unknown sample by both PGNAA and Fourier transform infrared spectroscopy (FT-IR) methods in which there were acceptable correlation between these two methods.

• Progress in Medical Physics
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• v.27 no.1
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• pp.31-36
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• 2016

The paper discusses radiation dose of dual energy CT on which copper modulation layer, is mounted in order to improve diagnostic performance of the dual energy CT. The radiation dose is estimated using MCNPX and its results are compared with that of the conventional dual energy CT system. CT X-ray spectra of 80 and 120 kVp, which are usually used for thorax, abdominal, head, and neck CT scans, were generated by the SPEC78 code and were used for the source specification 'SDEF' card for MCNPX dose modeling. The copper modulation layer was located 20 cm away from a source covering half of the X-ray window. The radiation dose was measured as changing its thickness from 0.5 to 2.0 mm at intervals of 0.5 mm. Since the MCNPX tally provides only normalized values to a single particle, the dose conversion coefficients of F6 tally for the modulation layer-based dual energy CBCT should be calculated for matching the modeling results into the actual dose. The dose conversion coefficient is $7.2*10^4cGy/output$ that is obtained from dose calibration curve between F6 tally and experimental results in which GAFCHORMIC EBT3 films were exposed by an already known source. Consequently, the dose of the modulation layer-based dual energy cone beam CT is 33~40% less than that of the single energy CT system. On the basis of the results, it is considered that scattered dose produced by the copper modulation layer is very small. It shows that the modulation layer-based dual energy CBCT system can effectively reduce radiation dose, which is the major disadvantage of established dual energy CT.

• Journal of the Korean Society of Radiology
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• v.11 no.7
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• pp.547-553
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• 2017

In the current medical field, lead is widely used as a radiation shield. However, the lead weight is very heavy, so wearing protective clothing such as apron is difficult to wear for long periods of time and there is a problem with the danger of lethal toxicity in humans. Recently, many studies have been conducted to develop substitute materials of lead to resolve these problems. As a substitute materials for lead, barium(Ba) and iodine(I) have excellent shielding ability. But, It has characteristics emitting characteristic X-rays from the energy area near 30 keV. For patients or radiation workers, shielding materials is often made into contact with the human body. Therefore, the characteristic X-rays generated by the shielding material are directly exposured in the human body, which increases the risk of increasing radiation absorbed dose. In this study, we have developed the FLUKA transport code, one of the most suitable elements of radiation transport codes, to remove the characteristic X-rays generated by barium or iodine. We have verified the reliability of the shielding fraction of the structure of the structure shielding by comparing with the MCPDX simulations conducted as a prior study. Using the MCNPX and FLUKA, the double layer shielding structures with the various thickness combination consisting of barium sulphate ($BaSO_4$) and bismuth oxide($Bi_2O_3$) are designed. The accuracy of the type shown in IEC 61331-1 was geometrically identical to the simulation. In addition, the transmission spectrum and absorbed dose of the shielding material for the successive x-rays of 120 kVp spectra were compared with lead. In results, $0.3mm-BaSO_4/0.3mm-Bi_2O_3$ and $0.1mm-BaSO_4/0.5mm-Bi_2O_3$ structures have been absorbed in both 33 keV and 37 keV characteristic X-rays. In addition, for high-energy X-rays greater than 90 keV, the shielding efficiency was shown close to lead. Also, the transport code of the FLUKA's photon transport code was showed cut-off on low-energy X-rays(below 33keV) and is limited to computerized X-rays of the low-energy X-rays. But, In high-energy areas above 40 keV, the relative error with MCNPX was found to be highly reliable within 6 %.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.4
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• pp.295-300
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• 2015

The precise determination of the activity for each radionuclide in environmental samples requires the self-absorption correction factor. In this research, we derived the self-absorption correction factor for three p-type high purity germanium detectors using the Monte Carlo code MCNPX. These detectors have different characteristics such as crystal diameter, height and size of the core. We compared the calculated full-energy peak efficiency with the experimental value using a standard sample with $1g/m^3$ density and verified the modeling. We simulated the dependency of the full-energy peak efficiency on the 0.3, 0.6, 0.9, 1.0, 1.2 and $1.5g/m^3$ samples and obtained the corresponding self-absorption correction factor. The self-absorption correction factors calculated for the three detectors differ by less than 1% over most of the energy range and sample densities considered. This indicates that the self-absorption correction factors are independent of the crystal characteristics of HPGe detector.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.151-158
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• 2018

The neutron powder diffractometer (NPD) is used to study a variety of technologically important and scientifically driven materials such as superconductors, multiferroics, catalysts, alloys, ceramics, cements, colossal magnetoresistance perovskites, magnets, thermoelectrics, zeolites, pharmaceuticals, etc. Monte Carlo-based codes are powerful tools to evaluate the neutronic behavior of the NPD. In the present study, MCNPX 2.6.0 and Vitess 3.3a codes were applied to simulate NPD facilities, which could be equipped with different optic devices such as pyrolytic graphite or neutron chopper. So, the Monte Carlo-based codes were used to simulate the NPD facility of the 5 MW Tehran Research Reactor. The simulation results were compared to the experimental data. The theoretical results showed good conformity to experimental data, which indicates acceptable performance of the Vitess 3.3a code in the neutron optic section of calculations. Another extracted result of this work shows that application of neutron chopper instead of monochromator could be efficient to keep neutron flux intensity higher than $10^6n/s/cm^2$ at sample position.

• Progress in Medical Physics
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• v.25 no.3
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• pp.151-156
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• 2014

The purpose of this study was to confirm the feasibility of imaging of therapy region from the boron neutron capture therapy (BNCT) using the measurement of the prompt gamma ray depending on the neutron flux. Through the Monte Carlo simulation, we performed the verification of physical phenomena from the BNCT; (1) the effects of neutron according to the existence of boron uptake region (BUR), (2) the internal and external measurement of prompt gamma ray dose, (3) the energy spectrum by the prompt gamma ray. All simulation results were deducted using the Monte Carlo n-particle extended (MCNPX, Ver.2.6.0, Los Alamos National Laboratory, Los Alamos, NM, USA) simulation tool. The virtual water phantom, thermal neutron source, and BURs were simulated using the MCNPX. The energy of the thermal neutron source was defined as below 1 eV with 2,000,000 n/sec flux. The prompt gamma ray was measured with the direction of beam path in the water phantom. The detector material was defined as the lutetium-yttrium oxyorthosilicate (Lu0,6Y1,4Si0,5:Ce; LYSO) scintillator with lead shielding for the collimation. The BUR's height was 5 cm with the 28 frames (bin: 0.18 cm) for the dose calculation. The neutron flux was decreased dramatically at the shallow region of BUR. In addition, the dose of prompt gamma ray was confirmed at the 9 cm depth from water surface, which is the start point of the BUR. In the energy spectrum, the prompt gamma ray peak of the 478 keV was appeared clearly with full width at half maximum (FWHM) of the 41 keV (energy resolution: 8.5%). In conclusion, the therapy region can be monitored by the gamma camera and single photon emission computed tomography (SPECT) using the measurement of the prompt gamma ray during the BNCT.

• Journal of the Korean Society of Radiology
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• v.11 no.3
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• pp.161-169
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• 2017

The cone beam computed tomography(CBCT) which can acquire 3-dimensions images is widely used for confirmation of patient position before radiation therapy. In this study, through the simulation using the Monte Carlo technique, we will analyze the exposure dose by cone beam computed tomography and present the standardized data. For the experiment, MCNPX(ver. 2.5.0) was used and the photon beam spectrum was analyzed after Cone beam was simulated. As a result of analyzing the photon beam spectrum, the average energy ranged from 25.7 to 37.6 keV at the tube voltage of 80 ~ 120 kVp and the characteristic X-ray energy was 9, 60, 68 and 70 keV. As a result of using the water phantom, the percentage depth dose was measured, and the maximum dose appeared on the surface and decreased with depth. The absorbed dose also decreased as the depth increased. The absorbed dose of the whole phantom was 9.7 ~ 18.7 mGy. This is a dose which accounts for 0.2% of about 10 Gy, which is generally used for radiation therapy per week, which is not expected to have a significant effect on the treatment effect. However, it should not be overlooked even if it is small compared with prescription dose.

• Journal of radiological science and technology
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• v.41 no.2
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• pp.141-148
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• 2018

This study examined the properties of photons and the dose distribution in a human body via a simulation where the total body irradiation(TBI) is performed on a pediatric anthropomorphic phantom and a child size water phantom. Based on this, we tried to find the optimal photon beam energy and material for beam spoiler. In this study, MCNPX (Ver. 2.5.0), a simulation program based on the Monte Carlo method, was used for the photon beam analysis and TBI simulation. Several different beam spoiler materials (plexiglass, copper, lead, aluminium) were used, and three different electron beam energies were used in the simulated accelerator to produce photon beams (6, 10, and 15 MeV). Moreover, both a water phantom for calculating the depth-dependent dosage and a pediatric anthropomorphic phantom for calculating the organ dosage were used. The homogeneity of photon beam was examined in different depths for the water phantom, which shows the 20%-40% difference for each material. Next, the org an doses on pediatric anthropomorphic phantom were examined, and the results showed that the average dose for each part of the body was skin 17.7 Gy, sexual gland 15.2 Gy, digestion 13.8 Gy, liver 11.8 Gy, kidney 9.2 Gy, lungs 6.2 Gy, and brain 4.6 Gy. Moreover, as for the organ doses according to materials, the highest dose was observed in lead while the lowest was observed in plexiglass. Plexiglass in current use is considered the most suitable material, and a 6 or 10 MV photon energy plan tailored to the patient condition is considered more suitable than a higher energy plan.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4671-4678
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• 2022

The purpose of this comprehensive research is to observe the impact of scintillator crystal type on entire detection process. For this aim, MCNPX (version 2.6.0) is used for designing of a physical plastic scintillation detector available in our laboratory. The modelled detector structure is validated using previous studies in the literature. Next, different types of plastic scintillation crystals were assessed in the same geometry. Several fundamental detector properties are determined for six different plastic scintillation crystals. Additionally, the deposited energy quantities were computed using the MCNPX code. Although six scintillation crystals have comparable compositions, the findings clearly indicate that the crystal composed of PVT 80% + PPO 20% has superior counting and detecting characteristics when compared to the other crystals investigated. Moreover, it is observed that the highest deposited energy amount, which is a result of the highest collision number in the crystal volume, corresponds to a PVT 80% + PPO 20% crystal. Despite the fact that plastic detector crystals have similar chemical structures, this study found that performing advanced Monte Carlo simulations on the detection discrepancies within the structures can aid in the development of the most effective spectroscopy procedures by ensuring maximum efficiency prior to and during use.

• Journal of the Korean Society of Radiology
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• v.11 no.7
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• pp.671-677
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• 2017

In this study, scattering factors affecting the quality of medical images were quantitatively analyzed and investigated. MCNPX simulation was conducted by using ANSI phantom, made of tissue equivalent materials, to calculate the scattering ratio occurred by the increase of the object thickness. Then, the result of the simulation was compared with the result of actual radiation measurement. In addition, we evaluated the image quality by the RMS evaluation, RSD and NPS analysis using X-ray images acquired with increasing object thickness. Furthermore, the scattering ratio was analyzed by increasing the thickness of acrylic phantom on chest phantom. The result showed that the scattering ratio was increased to 57.2%, 62.4%, and 66.8% from 48.9%, respectively, when the acrylic phantom thickness was increased by 1 inch from 6.1 inches. The results of MCNPX simulation and the actual measured scattering dose showed similar results. Also, as a result of RMS measurement from acquired x-ray images, the standard deviation decreased as the object thickness increased. However, in the RSD analysis considering the average incident dose, the results were increased from 0.028 to 0.039, 0.051, 0.062 as the acrylic phantom thickness was increased from 6.1 inches to 7.1 inch, 8.1 inch, and 9.1 inch, respectively. It can be seen that the increase of the scattering effect due to the increase of the object thickness reduces the SNR. Also, the NPS results obtained by measuring scattered radiation incident on the detector resulted in the increase of the noise as the object thickness increased.