• Journal of the Korean Society of Radiology
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• v.15 no.4
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• pp.415-427
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• 2021

Korea Institute of Radiological and Medical Sciences has provided various beam irradiation services to researchers using a 50 MeV cyclotron beam line. In particular, since the neutron beam service uses the nuclear reaction between protons and beryllium, the possibility of activation of the irradiated sample increases by using a high current. In this study, MCNP 6.2 and FISPACT-II 4.0 were used to evaluate the possible activation during the 35 MeV 20 ㎂ neutron beam service, which is preferred by the researchers. As a result of the calculation, if the iron, copper, and tungsten samples were irradiated for more than 1 hour, long-lived radioisotopes were produced and their radioactivity exceeded the standard level for self-disposal. Under the conditions of 2 hours of daily irradiation, no activation occurred in the building materials, and the internal exposure of workers due to air activation inside the irradiation room was very insignificant. And when this air was discharged to environment, the radioactivity including this air was also satisfied the emission standard.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2572-2578
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• 2022

An approximate method for calculating the cadmium ratio of a cyclotron-based thermal neutron radiography system was developed. In this method, the Monte-Carlo code, MCNP6.2, was employed to calculate the neutron capture rates of Au-197, and the cadmium ratio was obtained by computing the ratio of neutron capture rates. From the simulation results, the computed cadmium ratio is reasonably acceptable, and the assumption of ignoring the fast neutron contribution to the cadmium ratio is valid.

• Journal of Environmental Science International
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• v.13 no.2
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• pp.175-180
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• 2004

A prompt gamma-ray neutron activation (PGNA) system was simulated by the Monte Carlo N-Particle transport code (MCNP-4A) to estimate the level at which the scattered photon fluence rate, the absolute efficiency of the HPGe-detector, the volume of the concrete sample and the $^{35}$ /Cl(n, ${\gamma}$) reaction rate in this sample contribute to the count rate in the NaCl concentration measurement. The n- ${\gamma}$ fluence rates at the ST-2 beam tube exit of the HANARO reactor were used as input data, and the GAMMA-X type HPGe detector was modeled to tally 1.1649 MeV ${\gamma}$ -rays emitted from the $^{35}$ Cl(n, ${\gamma}$) reaction in the concrete sample. For three cylindrical concrete samples of 13.8, 46.8 and 157.1 ㎤ volumes, respectively, the relations between the NaCl weight fractions of 0.1, 1, 2 and 5 % in each of the concrete samples and the 1.1 649 MeV pulses created in the HPGe detector model were studied. As a result, it was found that the count rate at the same NaCl concentration nearly depends on the volume of the samples in a simulated condition of the same NaCl concentration samples, and that the linearities of the NaCl concentration calibration curves were reasonable in the narrow range of the NaCl weight fraction.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2221-2229
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• 2022

In this research, for improving the measurement performance of Prompt Gamma-ray Neutron Activation Analysis (PGNAA) set-up, a new optimization method for set-up design was proposed and investigated. At first, the calculation method for Signal-to-Noise Ratio (SNR) was proposed. Since the SNR could be calculated and quantified accurately, the SNR was chosen as the evaluation parameter in the new optimization method. For discussing the feasibility of the SNR optimization method, two kinds of PGNAA set-ups were designed in the MCNP code, based on the SNR optimization method and the previous signal optimization method, respectively. Meanwhile, the single element spectra analysis method was proposed, and the analysis effect of single element spectra as well as element sensitivity were used for comparing the measurement performance. Since the simulation results showed the better measurement performance of set-up designed by SNR optimization method, the experimental set-ups were built for the further testing, finally demonstrating the feasibility of the SNR optimization method for PGNAA setup design.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.337-343
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• 2021

In this work, self-absorption correction factor related to the variation of the composition and the density of soil samples were evaluated using the p-type HPGe detector. The validated MCNP5 simulation model of this detector was used to evaluate its Full Energy Peak Efficiency (FEPE) under the variation of the composition and the density of the analysed samples. The results indicates that FEPE calculation of low gamma ray is affected by the composition and the density of soil samples. The self-absorption correction factors for different gamma-ray energies which was fitted as a function of FEPEs via density and energy and fitting parameters as polynomial function for the logarithm neper of gamma ray energy help to calculate quickly the detection efficiency of detector. Factor Analysis for the influence of the element composition in analysed samples on the FEPE indicates the FEPE distribution changes from non-metal to metal groups when the gamma ray energy increases from 92 keV to 238 keV. At energies above 238 keV, the FEPE primarily depends only on the metal elements and is significantly affected by aluminium and silicon composition in soil samples.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.801-809
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• 2017

The purpose of this study is to assess the additional neutron effective dose during passive scattering proton therapy. Monte Carlo code (Monte Carlo N-Particle 6) simulation was conducted based on a precise modeling of the National Cancer Center's proton therapy facility. A three-dimensional neutron effective dose profile of the interior of the treatment room was acquired via a computer simulation of the 217.8-MeV proton beam. Measurements were taken with a $^3He$ neutron detector to support the simulation results, which were lower than the simulation results by 16% on average. The secondary photon dose was about 0.8% of the neutron dose. The dominant neutron source was deduced based on flux calculation. The secondary neutron effective dose per proton absorbed dose ranged from $4.942{\pm}0.031mSv/Gy$ at the end of the field to $0.324{\pm}0.006mSv/Gy$ at 150 cm in axial distance.

• Journal of Biosystems Engineering
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• v.38 no.2
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• pp.87-94
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• 2013

Purpose: Pineapple is now the third most important tropical fruit in world production after banana and citrus. Phytosanitary irradiation is recognized as a promising alternative treatment to chemical fumigation. However, most of the phytosanitary irradiation studies have dealt with physiochemical properties and its efficacy. Accurate dose calculation is crucial for ensuring proper process control in phytosanitary irradiation. The objective of this study was to optimize phytosanitary irradiation treatment of pineapple in various radiation sources using Monte Carlo simulation. Methods: 3-D geometry and component densities of the pineapple, extracted from CT scan data, were entered into a radiation transport Monte Carlo code (MCNP5) to obtain simulated dose distribution. Radiation energy used for simulation were 2 MeV (low-energy) and 10 MeV (high-energy) for electron beams, 1.25 MeV for gamma-rays, and 5 MeV for X-rays. Results: For low-energy electron beam simulation, electrons penetrated up to 0.75 cm from the pineapple skin, which is good for controlling insect eggs laid just below the fruit surface. For high-energy electron beam simulation, electrons penetrated up to 4.5 cm and the irradiation area occupied 60.2% of the whole area at single-side irradiation and 90.6% at double-side irradiation. For a single-side only gamma- and X-ray source simulation, the entire pineapple was irradiated and dose uniformity ratios (Dmax/Dmin) were 2.23 and 2.19, respectively. Even though both sources had all greater penetrating capability, the X-ray treatment is safer and the gamma-ray treatment is more widely used due to their availability. Conclusions: These results are invaluable for optimizing phytosanitary irradiation treatment planning of pineapple.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4447-4464
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• 2023

This paper focuses on the development of a new computational model of the CNESTEN's TRIGA Mark II research reactor using the 3D continuous energy Monte-Carlo code TRIPOLI-4 (T4). This new model was developed to assess neutronic simulations and determine quantities of interest such as kinetic parameters of the reactor, control rods worth, power peaking factors and neutron flux distributions. This model is also a key tool used to accurately design new experiments in the TRIGA reactor, to analyze these experiments and to carry out sensitivity and uncertainty studies. The geometry and materials data, as part of the MCNP reference model, were used to build the T4 model. In this regard, the differences between the two models are mainly due to mathematical approaches of both codes. Indeed, the study presented in this article is divided into two parts: the first part deals with the development and the validation of the T4 model. The results obtained with the T4 model were compared to the existing MCNP reference model and to the experimental results from the Final Safety Analysis Report (FSAR). Different core configurations were investigated via simulations to test the computational model reliability in predicting the physical parameters of the reactor. As a fairly good agreement among the results was deduced, it seems reasonable to assume that the T4 model can accurately reproduce the MCNP calculated values. The second part of this study is devoted to the sensitivity and uncertainty (S/U) studies that were carried out to quantify the nuclear data uncertainty in the multiplication factor k_eff. For that purpose, the T4 model was used to calculate the sensitivity profiles of the k_eff to the nuclear data. The integrated-sensitivities were compared to the results obtained from the previous works that were carried out with MCNP and SCALE-6.2 simulation tools and differences of less than 5% were obtained for most of these quantities except for the C-graphite sensitivities. Moreover, the nuclear data uncertainties in the k_eff were derived using the COMAC-V2.1 covariance matrices library and the calculated sensitivities. The results have shown that the total nuclear data uncertainty in the k_eff is around 585 pcm using the COMAC-V2.1. This study also demonstrates that the contribution of zirconium isotopes to the nuclear data uncertainty in the k_eff is not negligible and should be taken into account when performing S/U analysis.

• Journal of Sensor Science and Technology
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• v.25 no.3
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• pp.229-234
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• 2016

Self-powered neutron detector (SPND) is being widely used to monitor the reactor core of the nuclear power plants. The SPND contains a neutron-sensitive metallic emitter surrounded by a ceramic insulator. Currently, the vanadium (V) SPND has been being developed to be used in OPR1000 nuclear power plants. Some Monte Carlo simulations were accomplished to calculate the initial sensitivity of vanadium emitter material and alumina insulator with a cylindrical geometry. An MCNP code was used to simulate some factors (neutron self-shielding factor and beta escape probability from the emitter) and space charge effect of an insulator necessary to calculate the sensitivity of vanadium detector. The simulation results were compared with some theoretical and experimental values. The method presented here can be used to analyze the optimum design of the vanadium SPND and contribute to the development of TMI (Top-mount In-core Instrumentation) which might be used in the SMART and SMR.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2002.11a
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• pp.420-423
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• 2002

In recent years, there has been keen interest in developing f1at panel detectors for all modalities of radiology, including gerneral radiology, fluoroscopy(angiography and cardiology), electronic portal imaging, and mammography. In this paper, we report the new hybrid x-ray detector consisted of CsI(Tl) photoemission layer and a-Se photoconductor layer to resolve conventional x-ray detector such as the direct detector using a-Se and the indirect detector using CsI(Tl)/a-Si. To design the structure of CsI(Tl)/a-Se detector, the penetrated energy spectrum and absorption fraction was estimated using MCNP 4C code. Experimental results showed that the absorption fraction of $500{\mu}m-Se$ film and $150{\mu}m-CsI\left(Tl \right)/a-Se\left( 30{\mu}m \right)$ film is 70% at 70 kVp. The absorption energy is 90% at $350{\mu}m-CsI(Tl)$.