• Journal of the Korean Society of Safety
• /
• v.32 no.6
• /
• pp.29-33
• /
• 2017

In order to solve this sort of electrostatic failure in Display and Semiconductor process, Soft X-ray ionizer is mainly used. Soft X-ray Ionizer does not only generate electrical noise and minute particle but also is efficient to remove electrostatic as it has a wide range of ionization. There exist variable factors such as type of tungsten thickness deposited on target, Anode voltage etc., and it takes a lot of time and financial resource to find optimal performance by manufacturing with actual X-ray tube source. Here, MCNPX (Monte Carlo N-Particle Extended) is used for simulation to solve this kind of problem, and optimum efficiency of X-ray generation is anticipated. In this study, X-ray generation efficiency was compared according to target material thickness using MCNPX and actual X-ray tube source under the conditions that tube voltage is 5 keV, 10 keV, 15 keV and the target Material is Tungsten(W). At the result, In Tube voltage 5 keV and distance 100 mm, optimal target thickness is $0.05{\mu}m$ and fastest decay time appears + decay time 0.28 sec. - deacy time 0.30 sec. In Tube voltage 10keV and distance 100 mm, optimal target Thickness is $0.16{\mu}m$ and fastest decay time appears + decay time 0.13 sec. - deacy time 0.12 sec. In the tube voltage 15 keV and distance 100 mm, optimal target Thickness is $0.28{\mu}m$ and fastest decay time appears + decay time 0.04 sec. - deacy time 0.05 sec.

• Nuclear Engineering and Technology
• /
• v.48 no.5
• /
• pp.1120-1125
• /
• 2016

The neutron decay constant, ${\alpha}$, and effective delayed neutron fraction, ${\beta}_{eff}$, are important parameters for the control of the dynamic behavior of nuclear reactors. For the heavy water zero power reactor (HWZPR), this document describes the measurements of the neutron decay constant by noise analysis methods, including variance to mean (VTM) ratio and endogenous pulse source (EPS) methods. The measured ${\alpha}$ is successively used to determine the experimental value of the effective delayed neutron fraction as well. According to the experimental results, ${\beta}_{eff}$ of the HWZPR reactor under study is equal to 7.84e-3. This value is finally used to validate the calculation of the effective delayed neutron fraction by the Monte Carlo methods that are discussed in the document. Using the Monte Carlo N-Particle (MCNP)-4C code, a ${\beta}_{eff}$ value of 7.58e-3 was obtained for the reactor under study. Thus, the relative difference between the ${\beta}_{eff}$ values determined experimentally and by Monte Carlo methods was estimated to be < 4%.

• Nuclear Engineering and Technology
• /
• v.49 no.7
• /
• pp.1505-1512
• /
• 2017

Because accelerator-based boron neutron capture therapy (BNCT) systems are planned for use in hospitals, entry into the medical room should be controlled as hospitals are generally assumed to be public and safe places. In this paper, computational investigation of the medical staff effective dose during BNCT has been performed in different situations using Monte Carlo N-Particle (MCNP4C) code and two voxel based male phantoms. The results show that the medical staff effective dose is highly dependent on the position of the medical staff. The results also show that the maximum medical staff effective dose in an emergency situation in the presence of a patient is ${\sim}25.5{\mu}Sv/s$.

• Journal of the Korean Society of Radiology
• /
• v.16 no.5
• /
• pp.527-536
• /
• 2022

The radiation shielding characteristic of neutron shielding material has been studied as the preliminary study in order to design cosmic-ray shielding material. Specially, Soft Magnetic Material, known to be effective in EMP and radiation shielding, has been investigated to check if the material would be applicable to cosmic-ray shielding. In this work, thermal neutron shielding experiment was conducted and the Monte Carlo N-Particle(MCNP) was applied to employ skymap.dat, which is cosmic-ray data embedded in MCNP. As a result, polyethylene, borated polyethylene, and carbon nano tube, containing carbon or hydrogen, have been found to be effective in reduction of neutron flux below 20 MeV (including thermal, epithermal, evaporation). In contrast, the materials composed of iron such as SS316 and Soft Magnetic Material show a good shielding performance in the cascade energy range (above 20 MeV). Since Soft Magnetic Material is consisting of 13% of boron, it can also decrease thermal neutron flux, so it is expected that it would show a significant reduction on the entire range of neutron energy if the Soft Magnetic Material is used with hydrogen and carbon, so called low Z material.

• Journal of the Korean Society of Safety
• /
• v.32 no.2
• /
• pp.34-37
• /
• 2017

In recent emerging industry, Display field becomes bigger and bigger, and also semiconductor technology becomes high density integration. In Flat Panel Display, there is an issue that electrostatic phenomenon results in fine dust adsorption as electrostatic capacity increases due to bigger size. Destruction of high integrated circuit and pattern deterioration occur in semiconductor and this causes the problem of weakening of thermal resistance. In order to solve this sort of electrostatic failure in this process, Soft X-ray ionizer is mainly used. Soft X-ray Ionizer does not only generate electrical noise and minute particle but also is efficient to remove electrostatic as it has a wide range of ionization. X-ray Generating efficiency has an effect on soft X-ray Ionizer affects neutralizing performance. There exist variable factors such as type of anode, thickness, tube voltage etc., and it takes a lot of time and financial resource to find optimal performance by manufacturing with actual X-ray tube source. MCNPX (Monte Carlo N-Particle Extended) is used for simulation to solve this kind of problem, and optimum efficiency of X-ray generation is anticipated. In this study, X-ray generation efficiency was measured according to target material thickness using MCNPX under the conditions that tube voltage is 5 keV, 10 keV, 15 keV and the target Material is Tungsten(W), Gold(Au), Silver(Ag). At the result, Gold(Au) shows optimum efficiency. In Tube voltage 5 keV, optimal target thickness is $0.05{\mu}m$ and Largest energy of Light flux appears $2.22{\times}10^8$ x-ray flux. In Tube voltage 10 keV, optimal target Thickness is $0.18{\mu}m$ and Largest energy of Light flux appears $1.97{\times}10^9$ x-ray flux. In Tube voltage 15 keV, optimal target Thickness is $0.29{\mu}m$ and Largest energy of Light flux appears $4.59{\times}10^9$ x-ray flux.

• Journal of Radiation Protection and Research
• /
• v.41 no.2
• /
• pp.123-131
• /
• 2016

Background: With the increase in the number of particle accelerator facilities under either operation or construction, the accurate calculation using Monte Carlo codes become more important in the shielding design and radiation safety evaluation of accelerator facilities. Materials and Methods: The calculations with different physics models were applied in both of cases: using only physics model and using the mix and match method of MCNPX code. The issued conditions were the interactions of 600 MeV proton and $290MeV{\cdot}n^{-1}$ oxygen with a carbon target. Both of cross-section libraries, JENDL High Energy File 2007 (JENDL/HE-2007) and LA150, were tested in this calculation. In the case of oxygen ion interactions, the calculation results using LAQGSM physics model and JENDL/HE-2007 library were compared with D. Satoh's experimental data. Other Monte Carlo calculations using PHITS and FLUKA codes were also carried out for further benchmarking study. Results and Discussion: It was clearly found that the physics models, especially intra-nuclear cascade model, gave a great effect to determine proton-induced secondary neutron spectrum in MCNPX code. The variety of physics models related to heavy ion interactions did not make big difference on the secondary particle productions. Conclusion: The variations of secondary neutron spectra and particle transports depending on various physics models in MCNPX code were studied and the result of this study can be used for the shielding design and radiation safety evaluation.

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

• Nuclear Engineering and Technology
• /
• v.48 no.3
• /
• pp.624-634
• /
• 2016

Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that $k_{eff}$ values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

• Nuclear Engineering and Technology
• /
• v.53 no.3
• /
• pp.942-948
• /
• 2021

The aim of this work is to help inform the decision for choosing a convenient material for the PGAA (Prompt Gamma Activation Analysis) collimator plug to be installed at the tangential channel of the Moroccan Triga Mark II Research Reactor. Two families of materials are usually used for collimator construction: a mixture of high-density polyethylene (HDPE) with boron, which is commonly used to moderate and absorb neutrons, and heavy materials, either for gamma absorption or for fast neutron absorption. An investigation of two different collimator designs was performed using N-Particle Monte Carlo MCNP6.2 code with the ENDF/B-VII.1 and MCLIP84 libraries. For each design, carbon steel and lead materials were used separately as collimator heavy materials. The performed study focused on both the impact on neutron beam quality and the neutron-gamma background at the exit of the collimator beam tube. An analysis and assessment of the principal findings is presented in this paper, as well as recommendations.

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1885-1891
• /
• 2023

The present work aims to fabricate Na_1+xZr₂Si_xP_3-xO₁₂ compound at various calcination temperatures based on the zircon mineral. The fabricated compound was calcinated at 250, 500, and 1000℃. The effect of calcination temperature on the structure, crystal phase, and radiation shielding properties was studied for the fabricated compound. The X-ray diffraction diffractometer demonstrates that, the monoclinic crystal phase appeared at a calcination temperature of 250℃ and 500℃ is totally transformed to a high-symmetry hexagonal crystal phase under a calcination temperature of 1000℃. The radiation shielding capacity was also qualified for the fabricated compounds using the Monte Carlo N-Particle transport code in the g-photons energy interval between 15keV and 122keV. The impacts of calcination temperature on the g-ray shielding behavior were clarified in the present study, where the linear attenuation coefficient was enhanced by 218% at energy of 122keV, when the calcination temperature increased from 250 to 1000℃, respectively.