• Progress in Medical Physics
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• v.26 no.1
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• pp.12-17
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• 2015

The purpose of the Monte Carlo simulation study was to provide the optimized nozzle design to satisfy the beam conditions for biomedical researches in the Korean heavy-ion accelerator, RAON. The nozzle design was required to produce $C^{12}$ beam satisfying the three conditions; the maximum field size, the dose uniformity and the beam contamination. We employed the GEANT4 toolkit in Monte Carlo simulation to optimize the nozzle design. The beams for biomedical researches were required that the maximum field size should be more than $15{\times}15cm^2$, the dose uniformity was to be less than 3% and the level of beam contamination due to the scattered radiation from collimation systems was less than 5% of total dose. For the field size, we optimized the tilting angle of the circularly rotating beam controlled by a pair of dipole magnets at the most upstream of the user beam line unit and the thickness of the scatter plate located downstream of the dipole magnets. The values of beam scanning angle and the thickness of the scatter plate could be successfully optimized to be $0.5^{\circ}$ and 0.05 cm via this Monte Carlo simulation analysis. For the dose uniformity and the beam contamination, we introduced the new beam configuration technique by the combination of scanning and static beams. With the combination of a central static beam and a circularly rotating beam with the tilting angle of $0.5^{\circ}$ to beam axis, the dose uniformity could be established to be 1.1% in $15{\times}15cm^2$ sized maximum field. For the beam contamination, it was determined by the ratio of the absorbed doses delivered by $C^{12}$ ion and other particles. The level of the beam contamination could be achieved to be less than 2.5% of total dose in the region from 5 cm to 17 cm water equivalent depth in the combined beam configuration. Based on the results, we could establish the optimized nozzle design satisfying the beam conditions which were required for biomedical researches.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.1
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• pp.51-62
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• 2020

This study was conducted to develop a method for depth assessment of embedded sources using gamma-spectrum ratio and for the evaluation of field applicability. To this end, Peak to Compton and Peak to valley ratio changes were evaluated according to ¹³⁷Cs, ⁶⁰Co, ¹⁵²Eu point source depth using HPGe detector and MCNP simulation. The effects of measurement distance of PTV and PTC methods were evaluated. Using the results, the source depth assessment equation using the PTC and PTV methods was derived based on the detection distance of 50 cm. In addition, the sensitivity of detection distance changes was assessed when using PTV and PTC methods, and error increased by 3 to 4 cm when detection distance decreased by 20 cm based on 50 cm. However, it was confirmed that if the detection distance was increased to 100 cm, the effects of detection distance were small. And PTV and PTC methods were compared with the two distance measurement method which evaluates the depth of source by the change of net peak counting rate according to the detection distance. As a result of source depth assessment, the PTV and PTC showed a maximum error of 1.87 cm and the two distance measurement method showed maximum error of 2.69 cm. The results of the experiment confirmed that the accuracy of the PTV and PTC methods was higher than two distance measurement. In addition, Sensitivity evaluation by horizontal position error of source has maximum error of less than 25.59 cm for the two distance measurement method. On the other hand, PTV and PTC method showed high accuracy with maximum error of less than 8.04 cm. In addition, the PTC method has lowest standard deviation for the same time measurement, which is expected to enable rapid measurement.

• Journal of Radiation Protection and Research
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• v.39 no.4
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• pp.199-205
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• 2014

In the field of neutron detections, $^3He$ gas, the so-called "the gold standard," is the most widely used material for neutron detections because of its high efficiency in neutron capturing. However, from variable causes since early 2009, $^3He$ is being depleted, which has maintained an upward pressure on its cost. For this reason, the demands for $^3He$ replacements are rising sharply. Research into neutron converting materials, which has not been used well due to a neutron detection efficiency lower than the efficiency of $^3He$, although it can be chosen for use in a neutron detector, has been highlighted again. $^{10}B$, which is one of the $^3He$ replacements, such as $BF_3$, $^6Li$, $^{10}B$, $Gd_2O_2S$, is being researched by various detector development groups owing to a number of advantages such as easy gamma-ray discrimination, non-toxicity, low cost, etc. One of the possible techniques for the detection is an indirect neutron detection method measuring secondary radiation generated by interactions between neutrons and $^{10}B$. Because of the mean free path of alpha particle from interactions that are very short in a solid material, the thickness of $^{10}B$ should be thin. Therefore, to increase the neutron detection efficiency, it is important to make a $^{10}B$ thin film. In this study, we fabricated a $^{10}B$ thin film that is about 60 um in thickness for neutron detection using well-known technology for the manufacturing of a thin electrode for use in lithium ion batteries. In addition, by performing simple physical tests on the conductivity, dispersion, adhesion, and flexibility, we confirmed that the physical characteristics of the fabricated $^{10}B$ thin film are good. Using the fabricated $^{10}B$ thin film, we made a proportional counter for neutron monitoring and measured the neutron pulse height spectrum at a neutron facility at KAERI. Furthermore, we calculated using the Monte Carlo simulation the change of neutron detection efficiency according to the number of thin film layers. In conclusion, we suggest a fabrication method of a $^{10}B$ thin film using the technology used in making a thin electrode of lithium ion batteries and made the $^{10}B$ thin film for neutron detection using suggested method.

• Journal of the Korean Society of Radiology
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• v.15 no.6
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• pp.803-811
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• 2021

In this study, the exposure amount of IASCC test worker was evaluated by applying the process simulation technology. Using DELMIA Version 5, a commercial process simulation code, IASCC test facility, hot cells, and workers were prepared, and IASCC test activities were implemented, and the cumulative exposure of workers passing through the dose-distributed space could be evaluated through user coding. In order to simulate behavior of workers, human manikins with a degree of freedom of 200 or more imitating the human musculoskeletal system were applied. In order to calculate the worker's exposure, the coordinates, start time, and retention period for each posture were extracted by accessing the sub-information of the human manikin task, and the cumulative exposure was calculated by multiplying the spatial dose value by the posture retention time. The spatial dose for the exposure evaluation was calculated using MCNP6 Version 1.0, and the calculated spatial dose was embedded into the process simulation domain. As a result of comparing and analyzing the results of exposure evaluation by process simulation and typical exposure evaluation, the annual exposure to daily test work in the regular entrance was predicted at similar levels, 0.388 mSv/year and 1.334 mSv/year, respectively. Exposure assessment was also performed on special tasks performed in areas with high spatial doses, and tasks with high exposure could be easily identified, and work improvement plans could be derived intuitively through human manikin posture and spatial dose visualization of the tasks.