• Journal of the Semiconductor & Display Technology
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• v.9 no.4
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• pp.67-70
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• 2010

X-ray absorption rate and multiplication factor of X-ray detector were calculated with Garfield code. High Z (= atomic number) was important factor to increase the absorption rate but low Z is also important to increase the multiplication. Five different gas composition were examined under the condition of 1400 V bias and 400 um gap. Xe 100% gas showed the highest absorption rate and He 96% + isobutene 4% showed the highest multiplication.

• Wind and Structures
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• v.27 no.5
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• pp.283-291
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• 2018

This study presents the reliability assessment of a 100.5 m tall reinforced concrete chimney at a glass factory under wind loading by using vibration-based identified modal values. Ambient vibration measurements were recorded and modal values such as frequencies, shapes and damping ratios were identified by using Enhanced Frequency Domain Decomposition (EFDD) method. Afterwards, Finite Element Model (FEM) of the chimney was verified based on identified modal parameters. Reliability assessment of the chimney under wind loading was performed by obtaining the exceedance probability of demand to capacity distribution. Demand distribution of the chimney was developed under repetitive seeds of multivariate stochastic wind fields generated along the height of chimney. Capacity distribution of the chimney was developed by Monte Carlo simulation. Finally, it was found that reliability of the chimney is lower than code suggested limit values.

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.4
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• pp.351-359
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• 2009

We have developed a computer simulator fur cone-beam computed tomography (CBCT) based on the commercial Monte Carlo code, MCNP. All the functions to generate input files, run MCNP, convert output files to image data, reconstruct tomographs were realized in graphical user-interface form. The performance of the simulator was demonstrated by comparing with the experimental data. Although some discrepancies were observed due to the ignorance of the detailed physics in the simulation, such as scattered X-rays and noise in image sensors, the overall tendency was well agreed between the measured and simulated data. The developed simulator will be very useful for understanding the operation and the better design of CT systems.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1051-1059
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• 2018

Fast periodic pulsed reactor is a type of reactor in which the fission bursts are formed entirely with external reactivity modulation with a specified time periodicity. This type of reactors could generate much larger intensity of neutron beams for experimental use, compared with the steady state reactors. In the design of fast periodic pulsed reactors, the time dependent simulation of the power pulse is majorly based on a point kinetic model, which is known to have limitations. A more accurate calculation method is desired for the design analyses of fast periodic pulsed reactors. Monte Carlo computer code MCNP6 is used for this task due to its three dimensional transport capability with a continuous energy library. Some new routines were added to simulate the rotation of the movable reflector parts in the time dependent calculation. Fast periodic pulsed reactor IBR-2M was utilized to validate the new routines. This reactor is periodically in prompt supercritical state, which lasts for ${\sim}400{\mu}s$, during the equilibrium state. This generates long neutron fission chains, which requires tremendously large amount of computation time during Monte Carlo simulations. Russian Roulette was applied for these very long neutron chains in MCNP6 calculation, combined with other approaches to improve the efficiency of the simulations. In the power pulse of the IBR-2M at equilibrium state, there is some discrepancy between the experimental measurements and the calculated results using the point kinetics model. MCNP6 results matches better the experimental measurements, which shows the merit of using MCNP6 calculation relative to the point kinetics model.

• Journal of Radiation Protection and Research
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• v.48 no.1
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• pp.44-51
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• 2023

Background: Aptamers are currently being used in various fields including medical treatments due to their characteristics of selectively binding to specific molecules. Due to their special characteristics, the aptamers are expected to be used to remove radionuclides from a large amount of liquid radioactive waste generated during the decommissioning of nuclear power plants. The radiological effects on the aptamers should be evaluated to ensure their integrity for the application of a radionuclide removal technique. Materials and Methods: In this study, Monte Carlo N-Particle transport code version 6 (MCNP6) and Monte Carlo damage simulation (MCDS) codes were employed to evaluate the radiological effects on the aptamers. MCNP6 was used to evaluate the secondary electron spectrum and the absorbed dose in a medium. MCDS was used to calculate the DNA damage by using the secondary electron spectrum and the absorbed dose. Binding experiments were conducted to indirectly verify the results derived by MCNP6 and MCDS calculations. Results and Discussion: Damage yields of about 5.00×10^-4 were calculated for 100 bp aptamer due to the radiation dose of 1 Gy. In experiments with radioactive materials, the results that the removal rate of the radioactive ⁶⁰Co by the aptamer is the same with the non-radioactive ⁵⁹Co prove the accuracy of the previous DNA damage calculation. Conclusion: The evaluation results suggest that only very small fraction of significant number of the aptamers will be damaged by the radioactive materials in the liquid radioactive waste.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.301-308
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• 2024

Ceramic materials are being explored as alternatives to toxic lead sheets for radiation shielding due to their favorable properties like durability, thermal stability, and aesthetic appeal. However, crafting effective ceramics for radiation shielding entails complex processes, raising production costs. To investigate local viability, this study evaluated Malaysian ceramic tiles for shielding in diagnostic X-ray rooms. Different ceramics in terms of density and thickness were selected from local manufacturers. Energy Dispersive X-ray Fluorescence (EDXRF) and X-ray Fluorescence (XRF) characterized ceramic compositions, while Monte Carlo Particle and Heavy Ion Transport code System (MC PHITS) simulations determined Linear Attenuation Coefficient (LAC), Half-value Layer (HVL), Mass Attenuation Coefficient (MAC), and Mean Free Path (MFP) within the 40-150 kV energy range. Comparative analysis between MC PHITS simulations and real setups was conducted. The C3-S9 ceramic sample, known for homogeneous full-color structure, showcased superior shielding attributes, attributed to its high density and iron content. Notably, energy levels considerably impacted radiation penetration. Overall, C3-S9 demonstrated strong shielding performance, underlining Malaysia's potential ceramic tile resources for X-ray room radiation shielding.

• Journal of the Korea Institute of Military Science and Technology
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• v.11 no.5
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• pp.23-33
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• 2008

In this paper, we verify the performance of LDPC coded OFDM/DS system by Monte-Carlo simulation of BER on Eb/No. The simulation results show that LDPC coded OFDM/DS has a strong anti-jamming characteristic over pulse-noise jammer and partial-band noise jammer. The performance of LDPC coded OFDM/DS system is evaluated on both faded waveforms and non-faded waveforms. For non-faded waveforms, high coding gain is attained due to LDPC, even when waveforms have short PN sequence and JSR is only 5dB. Especially, the increase in the repeated number of LDPC decoding enhances coding gain. However, faded waveforms cannot achieve sufficient average effect when PN sequence is short. High coding gain of faded waveforms can be achieved by extending length of PN sequence. In addition, we compare LDPC coded OFDM/DS system with Convolutional coded OFDM/DS system. The simulation results illustrate that when LDPC coded OFDM/DS system with short PN sequence has sufficient average effects, the system shows lower BER than Convolutional coded OFDM/DS system with long PN sequence.

• Earthquakes and Structures
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• v.12 no.3
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• pp.359-374
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• 2017

The seismic events in Northern Italy, May 2012, have revealed the seismic vulnerability of typical Italian precast industrial buildings. The aim of this paper is to present a seismic fragility model for Italian RC precast buildings, to be used in earthquake loss estimation and seismic risk assessment by comparing two building typologies and three different codes: D.M. 3-03-1975, D.M. 16-01-1996 and current Italian building code that has been released in 2008. Based on geometric characteristics and design procedure applied, ten different building classes were identified. A Monte Carlo simulation was performed for each building class in order to generate the building stock used for the development of fragility curves trough analytical method. The probabilistic distributions of geometry were mainly obtained from data collected from 650 field surveys, while the material properties were deduced from the code in place at the time of construction or from expert opinion. The structures were modelled in 2D frameworks; since the past seismic events have identified the beam-column connection as the weakest element of precast buildings, two different modelling solutions were adopted to develop fragility curves: a simple model with post processing required to detect connection collapse and an innovative modelling solution able to reproduce the real behaviour of the connection during the analysis. Fragility curves were derived using both nonlinear static and dynamic analysis.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.647-653
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• 2020

Stainless steel is used commonly in nuclear applications for shielding radiation, so in this study, three different types of new stainless steel samples were designed and developed. New stainless steel compound ratios were determined by using Monte Carlo Simulation program Geant 4 code. In the sample production, iron (Fe), nickel (Ni), chromium (Cr), silicium (Si), sulphur (S), carbon (C), molybdenum (Mo), manganese (Mn), wolfram (W), rhenium (Re), titanium (Ti) and vanadium (V), powder materials were used with powder metallurgy method. Total macroscopic cross sections, mean free path and transmission number were calculated for the fast neutron radiation shielding by using (Geant 4) code. In addition to neutron shielding, the gamma absorption parameters such as mass attenuation coefficients (MACs) and half value layer (HVL) were calculated using Win-XCOM software. Sulfuric acid abrasion and compressive strength tests were carried out and all samples showed good resistance to acid wear and pressure force. The neutron equivalent dose was measured using an average 4.5 MeV energy fast neutron source. Results were compared to 316LN type stainless steel, which commonly used in shielding radiation. New stainless steel samples were found to absorb neutron better than 316LN stainless steel at both low and high temperatures.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.207-214
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• 2001

The KN-12 spent nuclear fuel (SNF) transport cask is designed for transportation of up to 12 assemblies and is in standby status for being licensed in accordance with Korea Atomic Energy Act. To evaluate radiation shielding and criticality safety of the KN-12 cask, each case of study was carried out using MCNP4B Code. MCNP code is verified by performing benchmark calculation for the KSC-4 SNF cask designed in 1989. As a result of radiation safety evaluation for the KN-12 cask, calculated dose rates always satisfied the standards at the cask surface, at 2m from the surface in normal transport condition, and at 1 m from the surface in hypothetical accident condition. Maximum dose rate was always arisen on the side of the cask. For normal transport condition, photons primarily contribute to dose rate between two kinds of released sources, neutrons and photons, from spent nuclear fuel but for hypothetical accident condition, contrary case was resulted. The level of calculated dose rate was 27.8% of the limit at the cask surface, 89.3% at 2 m from the cask surface, and 25.1% at 1 m from the cask surface. For criticality analysis, keff resulting from the criticality analysis considering the condition of optimum partial flooding with fresh water is 0.89708(0.00065. The results confirm the standards recommended by all regulations on radiation safety.