• Bulletin of the Korean Chemical Society
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• v.18 no.9
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• pp.981-985
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• 1997

13C, 15N, and 29Si NMR chemical shifts have been computed for selected X-substituted silatranes (X=Cl, F, H, CH3) using Gauge-Including Atomic Orbitals (GIAO) at the Hartree-Fock level of theory. The isotropic 13C chemical shifts are largely insensitive to substituent-induced structural changes. In this study, the isotropic 13C chemical shifts between 1-methyl- and 1-hydrogensilatranes by GIAO-SCF calculation at the HF/6-31G level are very similar. But the results of 1-chloro- and 1-fluorosilatranes are about 4 ppm different from the experimental values. In contrast, the isotropic 15N and 29Si chemical shifts and the chemical shielding tensors are quite sensitive to substituent-induced structural changes. These trends are consistent with those of the experiment. The isotropic 15N chemical shift demonstrates a very clear correlation with Si-N distance. But in case of 29Si the correlations are not as clean as for the 15N chemical shift; the calculated variation in the 29Si chemical shift is much larger.

• Journal of Radiation Protection and Research
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• v.47 no.1
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• pp.50-57
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• 2022

Background: There are several proton therapy facilities in operation or planned in Taiwan, and these facilities are anticipated to not only treat cancer but also provide beam services to the industry or academia. The simplified approach based on the Monte Carlo-based data sets (source terms and attenuation lengths) with the point-source line-of-sight approximation is friendly in the design stage of the proton therapy facilities because it is intuitive and easy to use. The purpose of this study is to expand the Monte Carlo-based data sets to allow the simplified approach to cover the application of proton beams more widely. Materials and Methods: In this work, the MCNP6 Monte Carlo code was used in three simulations to achieve the purpose, including the neutron yield calculation, Monte Carlo-based data sets generation, and dose assessment in simple cases to demonstrate the effectiveness of the generated data sets. Results and Discussion: The consistent comparison of the simplified approach and Monte Carlo simulation results show the effectiveness and advantage of applying the data set to a quick shielding design and conservative dose assessment for proton therapy facilities. Conclusion: This study has expanded the existing Monte Carlo-based data set to allow the simplified approach method to be used for dose assessment or shielding design for beam services in proton therapy facilities. It should be noted that the default model of the MCNP6 is no longer the Bertini model but the CEM (cascade-exciton model), therefore, the results of the simplified approach will be more conservative when it was used to do the double confirmation of the final shielding design.

• Journal of the Korean Society of Radiology
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• v.2 no.3
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• pp.33-35
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• 2008

In the neutron capture experiment and calculation, the neutron absorption and scattering are very important. Especially these effects are conspicuous in the resonance energy region and below the thermal energy region. In the present study, we obtained energy dependent neutron absorption ratios of natural rhodium in energy region from 0.003 to 100 eV by MCNP-4B Code. The coefficients for neutron absorption was calculated for several types of thickness. In the lower energy region, neutron absorption is larger than higher region, because of large capture cross section (1/v). Furthermore it seems very different neutron absorption in the large resonance energy region. These results are very useful to decide the thickness of sample and shielding materials.

• Journal of Welding and Joining
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• v.10 no.2
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• pp.32-42
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• 1992

A method of optimization of process parameters in Arc Welding has been discussed in this paper. The method of investigation is based on the numerical calculation of weld bead by a finite element method and non-linear optimization technique is applied to estimated the optimization process parameters from the numerical calculation. The common package program(ANSYS 4.4A) was used to obtain the process parameters for a thin plate arc welding (TIG, CO$_{2}$). The results on some test are satisfactory and the used method of this paper is a useful guide to the optimum welding condition.

• Journal of Radiation Protection and Research
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• v.26 no.2
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• pp.67-71
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• 2001

In order to shield the neutrons affecting the background of Low Level Gamma Ray Spectrometer, a neutron shielder was designed. The method used in this study for neutron shielding was the deceleration of fast neutrons by high density polyethylene(HDPE) and the absorption of those slowing-down neutrons by $B_4C$. The calculation results of neutron Interaction in HDPE using Monte Carlo simulation code MCNP4B showed that the thermal-neutron flux was maximum at 10 cm thickness of HDPE. The results also showed that 95% of the thermal neutrons were absorbed by 2 mm thickness of $B_4C$ absorber Consisted of 30 w% $B_4C$ and 70 w% polymer. The results of the Monte Carlo calculation were in good agreement with the experimental value obtained by a neutron shielding apparatus designed for this purpose.

• Journal of the Korean Society of Radiology
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• v.17 no.5
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• pp.633-640
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• 2023

With the recent development of static and dynamic modulated brachytherapy methods in brachytherapy, which use radiation shielding to modulate the dose distribution to deliver the dose, the amount of parameters and data required for dose calculation in inverse treatment planning and treatment plan optimization algorithms suitable for new directional beam intensity modulated brachytherapy is increasing. Although intensity-modulated brachytherapy enables accurate dose delivery of radiation, the increased amount of parameters and data increases the elapsed time required for dose calculation. In this study, a GPU-based CUDA-accelerated dose calculation algorithm was constructed to reduce the increase in dose calculation elapsed time. The acceleration of the calculation process was achieved by parallelizing the calculation of the system matrix of the volume of interest and the dose calculation. The developed algorithms were all performed in the same computing environment with an Intel (3.7 GHz, 6-core) CPU and a single NVIDIA GTX 1080ti graphics card, and the dose calculation time was evaluated by measuring only the dose calculation time, excluding the additional time required for loading data from disk and preprocessing operations. The results showed that the accelerated algorithm reduced the dose calculation time by about 30 times compared to the CPU-only calculation. The accelerated dose calculation algorithm can be expected to speed up treatment planning when new treatment plans need to be created to account for daily variations in applicator movement, such as in adaptive radiotherapy, or when dose calculation needs to account for changing parameters, such as in dynamically modulated brachytherapy.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.13 no.11
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• pp.5140-5147
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• 2012

It is difficult for systematic and flexible control reflecting regional characteristics with only public policies that control the landscape. Also, in the event that there is no preceding quantitative index calculation, it is impossible for the public society to come to an agreement. Therefore, the development of a shielding analysis simulation methodology that makes data processing modeling that can be interlinked with the urban information system is a very meaningful study. Thus, this study presents urban space shielding simulation technologies and quantitative analysis methodologies using 3D graphic engines and deduces the optimal design by integrating the data of the geographic information system (GIS) in order to suggest the potential as an analysis model that can be used in future urban information systems.

• 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.

• Journal of radiological science and technology
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• v.33 no.4
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• pp.363-367
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• 2010

In this paper we proposed an easy method to calculate the thickness of primary protective barrier for radiographic equipment. The concrete was selected for the shielding material. The area of protective barrier was divided into a controlled area and a noncontrolled area. For the computation of thickness, the data in NCRP Report 49 and 51 was used. For radiographic equipments whose maximum tubevoltages are 100 and 150 kVp, the thicknesses of concrete were calculated as a function of distance. From the calculated data, four analytical models were acquired by fitting an exponential decay function. From the equations acquired by this study, the thickness of primary protective barrier can be calculated approximately.

• Nuclear Engineering and Technology
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• v.25 no.4
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• pp.561-569
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• 1993

DOT4.2-QAD-CG coupling method was used to analyze the dose rates outside the side and the bottom shield system of CANDU 6 plant. The average dose rates at the main airlock and the new fuel loading area are approximately 6 $\mu$Sv/h as it is required. The calculated dose rates have a good agreement with the measurements at the operating CANDU 6 plant. The method used in this paper can be applied to the radiation shielding analysis of Wolsong 2, 3, and 4 CANDU 6 type plants which will be constructed in the near future.