• Journal of Radiation Protection and Research
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• v.7 no.1
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• pp.49-55
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• 1982

Thermally stimulated currents of polymers have some properties as radiation dosimetry, especially polymer could be made as a good dosimeter in biological fields because of tissue equivalent material. We experimented the radiation response of polymers and attempted to apply it in clinical use. Polymers have the properties of thermoluminescence and thermally stimulated currents which are due to several kinds of charged particles such as dipoles, electronic trapped charges and mobile ions. Several peaks are datected in the thermally stimulated currents in polyethylene under vias field V, by heating from room temperature to $100^{\circ}C$ shortly after irradiation. As V increases, both the peak temperature $T_m$ and the activation energy H decreases, while the peak current $I_m$ increases. We plotted the $T_m-V\;and\;I_m-V$ curves and calculated the electron trap depth with the recombination operative TSC theory and compared the peak TSC with radiation doses.

• The Journal of the Korea Contents Association
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• v.16 no.7
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• pp.451-456
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• 2016

To analyse the absorbed radiation dose of the visual organs (eyes, corneas, lenses) during a head CT scan, a with the purpose of radiation protection was designed. Afterwards, the reduction rate of radiation dose when using an eye-shielding was analyzed. The results showed that the higher the energy, the higher the absorbed dose of the eyes. Excluding the head, the organs with high dose were the eyes, corneas, and lenses, respectively. Furthermore, the dose reduction rate before and after shielding was between 38% and 55% for the eyes, and between 35% and 52% for the corneas. In the case of the lenses, when the front was shielded, the reduction rate was 51%, and when the front and the side were shielded simultaneously, the reduction rate was 67%.

• Journal of Radiation Protection and Research
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• v.42 no.4
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• pp.197-204
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• 2017

Background: X-ray imaging detectors for the nondestructive cargo container inspection using MeV-energy X-rays should accurately portray the internal structure of the irradiated container. Internal and external factors can cause noise, affecting image quality, and scattered radiation is the greatest source of noise. To obtain a high-performance transmission image, the influence of scattered radiation must be minimized, and this can be accomplished through several methods. The scatter rejection method using an anti-scatter grid is the preferred method to reduce the impact of scattered radiation. In this paper, we present an evaluation the characteristics of the signal and noise according to physical and material changes in the anti-scatter grid of the imaging detector used in cargo container scanners. Materials and Methods: We evaluated the characteristics of the signal and noise according to changes in the grid ratio and the material of the anti-scatter grid in an X-ray image detector using MCNP6. The grid was composed of iron, lead, or tungsten, and the grid ratio was set to 2.5, 12.5, 25, or 37.5. X-ray spectrum sources for simulation were generated by 6- and 9-MeV electron impacts on the tungsten target using MCNP6. The object in the simulation was designed using metallic material of various thicknesses inside the steel container. Using the results of the computational simulation, we calculated the change in the scatter-to-primary ratio and the signal-to-noise ratio improvement factor according to the grid ratio and the grid material, respectively. Results and Discussion: Changing the grid ratios of the anti-scatter grid and the grid material decreased the scatter linearly, affecting the signal-to-noise ratio. Conclusion: The grid ratio and material of the anti-scatter grid affected the response characteristics of a container scanner using high-energy X-rays, but to a minimal extent; thus, it may not be practically effective to incorporate anti-scatter grids into container scanners.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.666-672
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• 2024

In the present work, a new brick series based on the Vietnamese white clay minerals from the Bat Trang was fabricated to be applied in the radiation protection applications during the decommissioning of the nuclear power reactors. The bricks were constructed under various pressure rates varied from 7.61 MPa to 114.22 MPa. The influence of pressure rate on the physical and γ-ray shielding properties were investigated in the study. The experimental measurement for the material's density using the MH-300A density meter showed an enhancement in the prepared bricks' density by 22.5 % with increasing the applied pressure rate while the bricks' porosity reduced by 31.2 % when the pressure rate increased from 7.61 MPa to 114.22 MPa. The increase in the fabricated bricks density and the reduction in their porosities enhances the bricks' linear attenuation coefficients as measured by the NaI (Tl) detector along the energy range extended from 0.662 MeV to 1.332 MeV. The linear attenuation coefficient increased by 13.8 %, 17.6 %, 17.0 %, and 17.1 % at gamma ray energies of 0.662 MeV, 1.173 MeV, 1.252 MeV, and 1.332 MeV, respectively. The enhancement in the linear attenuation coefficient increases the bricks' radiation protection efficiency by 10.22 %, 14.48 %, 14.09 %, and 14.26 % at gamma ray energies of 0.662 MeV, 1.173 MeV, 1.252 MeV, and 1.332 MeV, respectively.

• Journal of radiological science and technology
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• v.34 no.2
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• pp.141-147
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• 2011

We developed an alternative radiation shielding material which is economical and has high protection efficiency. We validated the material in the form of sheet to make an apron. We increased the rate of barium and mixed tourmaline into silicon to improve the flexibility and protection rate of the sheet. The results showed that the shielding effect at low radiation energy is good enough with both 5 mm and 7 mm thickness. In the future, we will perform a quantitative evaluation of the reproducibility, volumetric efficiency, and porosity in mixing the ingredients.

• Journal of Radiation Protection and Research
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• v.34 no.1
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• pp.9-14
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• 2009

High density polyethylene (HDPE) is degraded due to a radiation-induced oxidation when it is used as a neutron moderator in a neutron counter for a nuclear material accounting of spent fuels. The HDPE exposed to the gamma-ray emitted from the fission products in a spent nuclear fuel results in a radiation-induced degradation which changes its original molecular structure to others. So a neutron moderating power variation of HDPE, irradiated by a gamma radiation, was investigated in this work. Five HDPE moderator structures were exposed to the gamma radiation emitted from a $^{60}Co$ source to a level of $10^5-10^9$ rad to compare their post-irradiation properties. As a result of the neutron measurement test with 5 irradiated HDPE structures and a neutron measuring system, it was confirmed that the neutron moderating power for the $10^5$ rad irradiated HDPE moderator revealed the largest decrease when the un-irradiated pure one was used as a reference. It implies that a neutron moderating power variation of HDPE is not directly proportional to the integrated gamma dose rate. To clarify the cause of these changes, some techniques such as a FTIR, an element analysis and a densitometry were employed. As a result of these analyses, it was confirmed that the molecular structure of the gamma irradiated HDPEs had partially changed to others, and the contents of hydrogen and oxygen had varied during the process of a radiation-induced degradation. The mechanism of these changes cannot be explained in detail at present, and thus need further study.

• Journal of Radiation Protection and Research
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• v.48 no.2
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• pp.100-106
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• 2023

Background: As breast tissue expanders consist of metallic materials in the needle guard and ferromagnetic injection port, irradiation can produce radioactivation. Materials and Methods: A CPX4 (Mentor Worldwide LLD) breast tissue expander was exposed using the Versa HD (Elekta) linear accelerator. Two photon energies of 6 and 10 MV-flattening filter free (FFF) beams with 5,000 monitor units (MU) were irradiated to identify the types of radiation. Furthermore, 300 MU with 10 MV-FFF beam was exposed to the CPX4 breast tissue expander by varying the machine dose rates (MDRs) 600, 1,200, and 2,200 MU/min. To assess the instantaneous dose rates (IDRs) solely from the CPX4, a tissue expander was placed outside the treatment room after beam irradiation, and a portable radioisotope identification device was used to identify the types of radiation and measure IDR. Results and Discussion: After 5,000 MU delivery to the CPX4 breast tissue expander, the energy spectrum whose peak energy of 511 keV was found with 10 MV-FFF, while there was no resultant one with 6 MV-FFF. The time of each measurement was 1 minute, and the mean IDRs from the 10 MV-FFF were 0.407, 0.231, and 0.180 μSv/hr for the three successive measurements. Following 10 MV-FFF beam irradiation with 300 MU indicated around the background level from the first measurement regardless of MDRs. Conclusion: As each institute room entry time protocol varies according to the working hours and occupational doses, we suggest an addition of 1 minute from the institutes' own room entry time protocol in patients with CPX4 tissue expander and the case of radiotherapy vaults equipped with a maximum energy of 10 MV photon beams.

• Journal of Radiation Protection and Research
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• v.49 no.1
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• pp.29-39
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• 2024

Background: The gamma emission tomography (GET) device has been reported a reliable technique to inspect partial defects within spent nuclear fuel (SNF) of pin-by-pin level. However, the existing GET devices have low accuracy owing to the high attenuation and scatter probability for SNF inspection condition. The purpose of this study is to design and optimize a Yonsei single-photon emission computed tomography version 2 (YSECT.v.2) for fast inspection of SNF in water storage by acquisition of high-quality tomographic images. Materials and Methods: Using Geant4 (Geant4 Collaboration) and DETECT-2000 (Glenn F. Knoll et al.) Monte Carlo simulation, the geometrical structure of the proposed device was determined and its performance was evaluated for the ¹³⁷Cs source in water. In a Geant4-based assessment, proposed device was compared with the International Atomic Energy Agency (IAEA)-authenticated device for the quality of tomographic images obtained for 12 fuel sources in a 14 × 14 Westinghouse-type fuel assembly. Results and Discussion: According to the results, the length, slit width, and septal width of the collimator were determined to be 65, 2.1, and 1.5 mm, respectively, and the material and length of the trapezoidal-shaped scintillator were determined to be gadolinium aluminum gallium garnet and 45 mm, respectively. Based on the results of performance comparison between the YSECT.v.2 and IAEA's device, the proposed device showed 200 times higher performance in gamma-detection sensitivity and similar source discrimination probability. Conclusion: In this study, we optimally designed the GET device for improving the SNF inspection accuracy and evaluated its performance. Our results show that the YSECT.v.2 device could be employed for SNF inspection.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.283-292
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• 2022

Advanced radiation applications have been widely used and extended to many fields. As a result of this fact, choosing an appropriate shielding material based on the radiation application has become vital. In this regard, the integration of elements into polymer composites has been investigated and contributed to the quantity and quality of radiation shielding materials. This study reports photon attenuation parameters and electromagnetic shielding effectiveness of a novel polymer composite prepared with a matrix reinforced with three different proportions (5, 10, and 15 wt%) of niobium content. Addition of Nb dopant improves both photon attenuation and electromagnetic shielding effectiveness for the investigated composites. Therefore, Nb(15%) polymer composite with highest concentration has been found to be the best absorber for ionizing and non-ionizing radiations. Consequently, the performed analyzes provide evidences that the prepared Nb-reinforced polymer composite could be effectively used as photon radiation attenuator and electromagnetic shielding material.

• Journal of Radiation Protection and Research
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• v.37 no.2
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• pp.70-74
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• 2012

Silicon carbide (SiC) is a promising material for neutron detection at harsh environments because of its capability to withstand strong radiation fields and high temperatures. Two PIN-type SiC semiconductor neutron detectors, which can be used for nuclear power plant (NPP) applications, such as in-core reactor neutron flux monitoring and measurement, were designed and fabricated. As a preliminary test, MCNPX simulations were performed to estimate reaction probabilities with respect to neutron energies. In the experiment, I-V curves were measured to confirm the diode characteristic of the detectors, and pulse height spectra were measured for neutron responses by using a $^{252}Cf$ neutron source at KRISS (Korea Research Institute of Standards and Science), and a Tandem accelerator at KIGAM (Korea Institute of Geoscience and Mineral Resources). The neutron counts of the detector were linearly increased as the incident neutron flux got larger.