• Journal of Radiation Protection and Research
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• v.36 no.4
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• pp.200-208
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• 2011

The neutron fluence measurement and evaluation technology is very important for material irradiation test. The most essential technology in this study is the neutron irradiation evaluation method using a fluence monitor. The fluence monitors were fabricated with metal wires of the purity ${\geq}$ 99.9%, whose dimensions were 0.1mm diameter, about 3 mm length, and around 150-200 ${\mu}g$ mass range. Three wire samples (Fe, Ni, Ti) were prepared for one irradiation aluminum capsule. Five capsules were irradiated in the OR5 hole of the HANARO reactor at 30 MW power for about 25 days. After irradiation tests, radiation activities were measured with the high purity germanium (HPGe) detector. The reaction rates were calculated by using the measured radiation activity data, and then neutron fluence were obtained from the reaction rates and the weighted neutron cross section with calculated neutron spectrum at the fluence monitor position.

• 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.17 no.4
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• pp.489-495
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• 2023

Radiation shielding facilities are constructed in locations where diagnostic radiation generators are installed, with the aim of preventing exposure for patients and radiation workers. The purpose of this study is seek to compare and validate the trend of attenuation thickness of lead, the primary material in these radiation shielding facilities, at different maximum tube voltages by Monte Carlo simulations and measurement. We employed the Monte Carlo N-Particle 6 simulation code. Within this simulation, we set a lead shielding arrangement, where the distance between the source and the lead sheet was set at 100 cm and the field of view was set at 10 × 10 cm². Additionally, we varied the tube voltages to encompass 80, 100, 120, and 140 kVp. We calculated energy spectra for each respective tube voltage and applied them in the simulations. Lead thicknesses corresponding to attenuation rates of 50, 70, 90, and 95% were determined for tube voltages of 80, 100, 120, and 140 kVp. For 80 kVp, the calculated thicknesses for these attenuation rates were 0.03, 0.08, 0.21, and 0.33 mm, respectively. For 100 kVp, the values were 0.05, 0.12, 0.30, and 0.50 mm. Similarly, for 120 kVp, they were 0.06, 0.14, 0.38, and 0.56 mm. Lastly, at 140 kVp, the corresponding thicknesses were 0.08, 0.16, 0.42, and 0.61 mm. Measurements were conducted to validate the calculated lead thicknesses. The radiation generator employed was the GE Healthcare Discovery XR 656, and the dosimeter used was the IBA MagicMax. The experimental results showed that at 80 kVp, the attenuation rates for different thicknesses were 43.56, 70.33, 89.85, and 93.05%, respectively. Similarly, at 100 kVp, the rates were 52.49, 72.26, 86.31, and 92.17%. For 120 kVp, the attenuation rates were 48.26, 71.18, 87.30, and 91.56%. Lastly, at 140 kVp, they were measured 50.45, 68.75, 89.95, and 91.65%. Upon comparing the simulation and experimental results, it was confirmed that the differences between the two values were within an average of approximately 3%. These research findings serve to validate the reliability of Monte Carlo simulations and could be employed as fundamental data for future radiation shielding facility construction.