• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.792-800
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• 2017

The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma $({\gamma})-rays$. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM). The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, ${\gamma}-ray$ shielding behaviors (GSB) of six glass samples (transparent NEM) were evaluated and compared with some opaque NEM in a wide range of energy (15 keV-15 MeV) and optical thickness (OT). The study was performed by computing various ${\gamma}-ray$ shielding parameters (GSP) such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well.

• 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 Radiation Protection and Research
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• v.43 no.4
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• pp.154-159
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• 2018

Background: To prevent small leakage accidents, a real-time and direct detection system for small leaks with a detection limit below that of existing systems, e.g. $0.5gpm{\cdot}hr^{-1}$, is required. In this study, a small-size beta detector, which can be installed inside the reactor containment (CT) building and detect small leaks directly, was suggested and its feasibility was evaluated using MCNPX simulation. Materials and Methods: A target nuclide was selected through analysis of radiation from radionuclides in the reactor coolant system (RCS) and the spectrum was obtained via a silicon detector simulated in MCNPX. A window was designed to reduce the background signal caused by other nuclides. The sensitivity of the detector was also estimated, and its shielding designed for installation inside the reactor CT. Results and Discussion: The beta and gamma spectrum of the silicon detector showed a negligible gamma signal but it also contained an undesired peak at 0.22 MeV due to other nuclides, not the $^{16}N$ target nuclide. Window to remove the peak was derived as 0.4 mm for beryllium. The sensitivity of silicon beta detector with a beryllium window of 1.7 mm thickness was derived as $5.172{\times}10^{-6}{\mu}Ci{\cdot}cc^{-1}$. In addition, the specification of the shielding was evaluated through simulations, and the results showed that the integrity of the silicon detector can be maintained with lead shielding of 3 cm (<15 kg). This is a very small amount compared to the specifications of the lead shielding (600 kg) required for installation of $^{16}N$ gamma detector in inside reactor CT, it was determined that beta detector would have a distinct advantage in terms of miniaturization. Conclusion: The feasibility of the beta detector was evaluated for installation inside the reactor CT to detect small leaks below $0.5gpm{\cdot}hr^{-1}$. In future, the design will be optimized on specific data.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4130-4136
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• 2021

By polymerization of gadolinium methacrylate (Gd (MAA)₃), lead methacrylate (Pb(MAA)₂) and methyl methacrylate (MMA), Gd and Pb were chemically bonded into polymers. The X-ray shielding performance was evaluated by Monte Carlo simulation method, and the results showed that the more metal functional organic monomer, the better the shielding performance of terpolymers. When the X-ray energy is 65 keV, Gd (MAA)₃-containing polymers have better shielding performance than Pb(MAA)₂-containing polymers. Gd could compensate for the weak absorption region of Pb. Therefore, polymers containing both Gd and Pb enhanced shielding efficiency against X-ray in various low-energy ranges. For obtaining terpolymers with uniform monomer compositions, the relationship between the monomer composition of the terpolymers and the conversion level was optimized by calculating the reactivity ratios. The value of reactivity ratios of r (Gd (MAA)₃/Pb(MAA)₂), r (Pb(MAA)₂/Gd (MAA)₃), r (Gd (MAA)₃/MMA), r (MMA/Gd (MAA)₃), r (Pb(MAA)₂/MMA) and r (MMA/Pb(MAA)₂) was 0.483, 0.004, 0.338, 2.508, 0.255, 0.029. The terpolymers with uniform monomer composition could be obtained by controlling the monomer compositions or conversion levels. The results can provide new radiation protection materials and contribute to the improvement in nuclear safety.

• Journal of the Korean Society of Radiology
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• v.12 no.4
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• pp.467-474
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• 2018

CBCT is useful for improving the accuracy of the treatment site, but Repeated use increases the exposure dose. In this study, we aimed to provide basic data for dose reduction in CBCT implementation by dataization the simulating and dose reduction effect using shielding substance. Material in this study, Analyzation the photon beam by simulate the CBCT Through MCNPX and then calculate the absorption dose of body organ at shooting moment of thoracic abdominal position as target UF-Revise simulated body. At this time. Dose reduction effects at this time were evaluated according to the texture of materials and presence of shielding materials( lead, antimony, barium, sulfate, tungsten, bismuth). When CBCT was taken without shielding, the dose was calculated to be high in the breast and spine, and the dose in the esophagus and lung was calculated to be low. The doses according to the shield material were calculated as barium sulfate, antimony, bismuth, lead, and tungsten. The shielding rate was the highest in the thymus (73.6%) and the breast (59.9%) compared with the dose reduction according to presence or absence of the shield. However, it showed the lowest shielding rate in lung (2.1%) and spine (12.6%).

• Journal of the Korean Society of Radiology
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• v.14 no.5
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• pp.627-634
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• 2020

Instead of lead, we intend to develop shielding materials for morphological implementation by mixing barium sulfate, an eco-friendly substance, and PLA filament, a 3D printer material. The environmental substance, barium sulfate powder and PLA filament, a 3D printer material, were used, and the shielding was made with a 3D printer after being fused into an extruder to mix the powder powder of barium sulfate with PLA. To check the mixing ratio of barium sulfate powder and PLA filament, the mixing input was analyzed, and the absorption dose by thickness according to barium sulfate content was obtained to check the shielding function of the mixed shielding. In the evaluation of the mixture of sulfate barium powder particles and PLA filaments, it was mixed in the most appropriate proportion when the content was 30% in the apparent and electron microscopic observation photographs. In the absorption dose results by thickness according to barium sulfate content, the difference between the content of 0% and the content of each % was greatest at 0.5 cm in thickness and the lowest dose value at 3 cm in thickness when the barium content was 30%. In addition, as the barium content began to increase at 30%, the absorbed dose value increased again. Instead of conventional lead, barium sulfate, an eco-friendly substance, could be mixed with PLA, a filament material, to create morphological shielding. Based on this study, it is expected that the mixing ratio of barium to the mixture is the most appropriate 30%, and will be used as the basis for the implementation of morphological shielding using 3D printers in the diagnosis and treatment section.

• 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.

• The Journal of Korean Society for Radiation Therapy
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• v.31 no.1
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• pp.57-65
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• 2019

Purpose: The purpose is to clarify the effect of additional scattering ratio on the edge of the block according to the increasing block thickness with low melting point lead alloy and pure lead in electron beam therapy. Methods and materials: $10{\times}10cm^2$ Shielding blocks made of low melting point lead alloy and pure lead were fabricated to shield mold frame half of applicator. Block thickness was 3, 5, 10, 15, 20 (mm) for each material. The common irradiation conditions were set at 6 MeV energy, 300 MU / Min dose rate, gantry angle of $0^{\circ}$, and dose of 100 MU. The relative scattering ratio with increasing block thickness was measured with a parallel plate type ion chamber(Exradin P11) and phantom(RW3) by varying the position of the shielding block(cone and on the phantom), the position of the measuring point(surface ans depth of $D_{max}$), and the block material(lead alloy and pure lead). Results : When (depth of measurement / block position / block material) was (surface / applicator / pure lead), the relative value(scattering ratio) was 15.33 nC(+0.33 %), 15.28 nC(0 %), 15.08 nC(-1.31 %), 15.05 nC(-1.51 %), 15.07 nC(-1.37 %) as the block thickness increased in order of 3, 5, 10, 15, 20 (mm) respectively. When it was (surface / applicator / alloy lead), the relative value(scattering ratio) was 15.19 nC(-0.59 %), 15.25 nC(-0.20 %), 15.15 nC(-0.85 %), 14.96 nC(-2.09 %), 15.15 nC(-0.85 %) respectively. When it was (surface / phantom / pure lead), the relative value(scattering ratio) was 15.62 nC(+2.23 %), 15.59 nC(+2.03 %), 15.53 nC(+1.67 %), 15.48 nC(+1.31 %), 15.34 nC(+0.39 %) respectively. When it was (surface / phantom / alloy lead), the relative value(scattering ratio) was 15.56 nC(+1.83 %), 15.55 nC(+1.77 %), 15.51 nC(+1.51 %), 15.42 nC(+0.92 %), 15.39 nC(+0.72 %) respectively. When it was (depth of $D_{max}$ / applicator / pure lead), the relative value(scattering ratio) was 16.70 nC(-10.87 %), 16.84 nC(-10.12 %), 16.72 nC(-10.78 %), 16.88 nC(-9.93 %), 16.90 nC(-9.82 %) respectively. When it was (depth of $D_{max}$ / applicator / alloy lead), the relative value(scattering ratio) was 16.83 nC(-10.19 %), 17.12 nC(-8.64 %), 16.89 nC(-9.87 %), 16.77 nC(-10.51 %), 16.52 nC(-11.85 %) respectively. When it was (depth of $D_{max}$ / phantom / pure lead), the relative value(scattering ratio) was 17.41 nC(-7.10 %), 17.45 nC(-6.88 %), 17.34 nC(-7.47 %), 17.42 nC(-7.04 %), 17.25 nC(-7.95 %) respectively. When it was (depth of $D_{max}$ / phantom / alloy lead), the relative value(scattering ratio) was 17.45 nC(-6.88 %), 17.44 nC(-6.94 %), 17.47 nC(-6.78 %), 17.43 nC(-6.99 %), 17.35 nC(-7.42 %) respectively. Conclusions: When performing electron therapy using a shielding block, the block position should be inserted applicator rather than the patient's body surface. The block thickness should be made to the minimum appropriate shielding thickness of each corresponding using energy. Also it is useful that the treatment should be performed considering the influence of scattering dose varying with distance from the edge of block.

• Journal of the Korean Society of Radiology
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• v.11 no.5
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• pp.329-334
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• 2017

The purpose of this study is to estimating the possibility of manufacturing radiation shielding sheet by searching for environmentally friendly materials suitable for medical environment of medical radiation shielding. There are many tungsten products which are currently used as shielding materials in place of lead, but there are small problems in the mass production of lightweight shielding sheets due to economical efficiency. To solve these problems, a lightweight, environmentally friendly material with economical efficiency is required. In this study, Barium sulphate and Iodine were proposed. Both materials are already used as contrast medias in radiography, and it is predicted that the shielding effect will be sufficient in a certain region as a shielding material because of the characteristic of absorbing radiation. Therefore, in this study, we used a Monte Carlo simulation to simulate radiation shielding materials. When it is a contrast agent such as Barium sulfate and Iodine, the radiation absorption effect in the high energy region appears greatly, and the effectiveness of the two shielding substance in the energy region of the star with thickness of 120 kVp is also evaluated in the medical radiation imaging region. Simulated estimation results it was possible to estimate the effectiveness of shielding for all two substances. Iodine has higher shielding effect than barium sulfate, 0.05 mm thick appears great effect. Therefore, the Monte Carlo simulation confirms that iodine, which is a radiological contrast agent, is also usable as barium sulfate in the production of radiation shielding sheets.

• Journal of the Korea Convergence Society
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• v.10 no.12
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• pp.129-134
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• 2019

PVC was chosen as a plastic product that can cope with lead, a radiation shielding material that is widely used in medical institutions. In addition to radiation shielding clothing, we want to evaluate whether it can be used as a medical device component and industrial shielding material in low dose areas. Commercial PVC has a density of 3.68 g/㎠ and can be positively expected sufficient shielding effect in certain radiation areas such as material flexibility and economy efficiency, and can be transformed into various forms and used as a lightweight shielding wall. The shielding performance was tested by adjusting the thickness of 5 sheets of 3mm PVC in the range of medical radiation used for clinical examination in medical institutions. Shielding performance against effective energy was evaluated based on tube radiation voltage of medical radiation. The thicker the PVC, the lower the tube voltage and the lower the effective energy, the greater the shielding effect. The shielding effect was 70% at 12mm thickness and 80kVp tube voltage. Therefore, the shielding effect of PVC material has a high dependence of thickness. In the future, continuous research is needed to make thin and light eco-friendly products while improving shielding performance.