• Journal of Radiation Protection and Research
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• v.40 no.3
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• pp.174-180
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• 2015

There are currently many research papers on the knowledge, perceptions and actions of radiation-related staff, but hardly any papers on radiation major students in college who are to be staff members of radiation related jobs in the future. It is of course important to understand the perceptions of staff working on the lines and change their knowledge and perceptions, but in the long term it seems more efficient to understand those who are in the stage of being educated to be staff members-their knowledge and perception of radiation so that ultimately they can attain the right kind of understanding. Therefore the aim of this study is to grasp the pre-radiation staff's basic concept of radiation and space dose, their understanding of radiation safety based on this, and whether there is a change in their perception before and after the space dose measurement experiment; in the end this is to see if the space dose measurement experiment is effective in changing perception on radiation safety. This study took as its subject 64 students majoring radiation in college, I.e. pre-radiation staff members, and gauged their basic conceptualization of radiation, understanding of space dose, and understanding of radiation safety; in the X-ray room within the department the students were asked to measure space dose for themselves, so as to see whether there was a change in their understanding of radiation safety before and after the experiment, according to their understanding of the basic concept of radiation and of space dose. As a result of the space dose measurement experiment, students' increased basic knowledge of concept of radiation and understanding of the dangers of space dose were noteworthy, and accordingly their understanding of radiation safety became stricter and more conservative. In spite of this, their work ethic stayed in the lead of their understanding of radiation safety; this implies the need of a more departmentalized safety education program. Therefore instead of safety education that simply uses visual-audial material in a kind of lecture, I suggest here that there be a more experiential safety education program that enables learners to try out space dose measurement experiments for themselves, a work ethic education that aims for a conventional point of view towards radiation safety as well as a stern attitude.

• 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.

• Journal of Radiation Protection and Research
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• v.39 no.1
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• pp.7-13
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• 2014

Phosphate rock, phosphogypsum, and products in phosphate processing facility contain naturally occurring radioactive materials (NORM). Therefore, they may give rise to enhanced radiation dose to workers due to inhalation of airborne particulates. Internal dose due to particle inhalation varies depending on particle properties. The objective of the present study was to characterize particle properties at the largest phosphate processing facility in Korea. A cascade impactor was employed to sample airborne particulates at various processing areas in the plant. The collected samples were used for characterization of particle size distribution, particle concentration in the air, and shape analysis. Aerodynamic diameters of airborne particulates ranged 0.03-100 ${\mu}m$ with the highest concentration at the particle size range of 4.7-5.8 ${\mu}m$ (geometric mean = 5.22 ${\mu}m$) or 5.8-9.0 ${\mu}m$ (geometric mean = 7.22 ${\mu}m$). Particle concentrations in the air varied widely by sampling area up to more than two orders of magnitude. The large variation resulted from the variability of mechanical operations and building ventilations. The airborne particulates appeared as spheroids or rough spherical fragments across all sampling areas and sampled size intervals. Average mass densities of phosphate rocks, phosphogypsums, and fertilizers were 3.1-3.4, 2.1-2.6, and 1.7 $gcm^{-3}$, respectively. Radioactivity concentration of uranium series in phosphate rocks varied with country of origin, ranging 94-866 $Bqkg^{-1}$. Among the uranium series, uranium was mostly concentrated on products, including phosphoric acid or fertilizers whereas radium was concentrated on byproducts or phosphogypsum. No significant radioactivity of $^{226}Ra$ and $^{228}Ra$ were found in fertilizer. However, $^{40}K$ concentration in fertilizer was up to 5,000 Bq $g^{-1}$. The database established in this study can be used for the accurate risk assessment of workers due to inhalation of airborne particles containing NORM. In addition, the findings can be used as a basic data for development of safety standard and guide and for practical radiation safety management at the facility.

• Journal of radiological science and technology
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• v.40 no.1
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• pp.41-47
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• 2017

Considering that the X-ray apron used in the department of radiology is also used in the department of nuclear medicine, the study aimed to analyze the shielding rate of the apron according to types of radioisotopes, thus ${\gamma}$ ray energy, to investigate the protective effects. The radioisotopes used in the experiment were the top 5 nuclides in usage statistics $^{99m}Tc$, $^{18}F$, $^{131}I$, $^{123}I$, and $^{201}Tl$, and the aprons were lead equivalent 0.35 mmPb aprons currently under use in the department of nuclear medicine. As a result of experiments, average shielding rates of aprons were $^{99m}Tc$ 31.59%, $^{201}Tl$ 68.42%, and $^{123}I$ 76.63%. When using an apron, the shielding rate of $^{131}I$ actually resulted in average dose rate increase of 33.72%, and $^{18}F$ showed an average shielding rate of -0.315%, showing there was almost no shielding effect. As a result, the radioisotopes with higher shielding rate of apron was in the descending order of $^{123}I$, $^{201}Tl$, $^{99m}Tc$, $^{18}F$, $^{131}I$. Currently, aprons used in the nuclear medicine laboratory are general X-ray aprons, and it is thought that it is not appropriate for nuclear medicine environment that utilizes ${\gamma}$ rays. Therefore, development of nuclear medicine exclusive aprons suitable for the characteristics of radioisotopes is required in consideration of effective radiation protection and work efficiency of radiation workers.

• Journal of Preventive Medicine and Public Health
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• v.29 no.1 s.52
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• pp.1-14
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• 1996

To evaluate the health risk of the workplace environment of a general hospital toward the hospital workers, a questionnaire survey on the perception of risk at the workplace environment and environmental measurements at 27 locations with 240 workers in the hospital were made from October 25th, 1993 to October 30th, 1994. The results were as follows ; 1. By the environmental measurements, 86 workers(36%) were found to be exposed to poor or harmful degree of general illumination, 193 workers(80%) were exposed to poor or harmful local illumination, 34 workers(14%) were exposed to poor or harmful degree of thermal condition and 180 workers(75%) were exposed to poor or harmful noise level, but nobody was exposed to poor or harmful dust and toluene concentration. Also nobody was exposed to poor or harmful level of radiation or anticancer/antibiotic agents. 2. The subjective perceptions on the environmental conditions felt by the workers were different from the objective findings by the environmental measurements. The workers underscored the poor illumination state and overscored the dust concentrations. Also workers oversensitized about the thermal condition, the noise level, the radiation level, the toluene concentration and anticancer/antibiotic agents 3. The sources of noise were the dialogue(179 workers) and the office instruments(131 workers). The sources of dust came from the clothes(108 workers) and the building materials(79 workers). 4. The questionnaire survey showed that the 28 workers were exposed to poor or harmful level of the antibiotics, 10 workers to alkali or acid and 6 workers to drug dust in the pharmacy but the above findings could not be proved by the environmental measurements.

• Journal of the Korean Society of Radiology
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• v.12 no.1
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• pp.31-37
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• 2018

In the surface contamination test using the end-window Geiger-Muller type counter, the wrap is used as a method for protecting the detector exposed to the outside in order to measure the beta-rays. We analyze the effect of this method on the measurement rate and the correction factor, and wanted to make it clear to radiation workers that excessive use of the wrap can affect the measured value of the beta-rays. The experimental method was to compare and analyze the change of the beta-rays measurement counting rate and the calibration factor according to the wrap thickness using the beta-rays with different energy of 3 KBq, 1.5 KBq and 0.3 KBq. The subjects of this study were the end-window Geiger-Muller type counter which were held at the calibration center certified by Korea Laboratory Accreditation Scheme (KOLAS) in March 2012, Cl-36 (Chlorine) and Sr-90 (Strontium) were used as the source of beta radiation. The measurement counting rate decreased with increasing wrap thickness, and the calibration factor increased with increasing wrap thickness. Since the changes of the measurement counting rate and the calibration factors can reduce the accuracy of the instrument readings, but also have a significant impact on detector contamination and damage, so there is a need to find out what thickness of wrap is most effective. If we using a wraps with thickness that show a low rate of change of the measurement counting rate and the calibration factor, it will protect the detector and minimize the effect on the measured value of the beta-rays.

• Journal of Radiation Protection and Research
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• v.37 no.2
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• pp.103-107
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• 2012

This paper analyzes changes in the external radiation dose rate of PET-CT test patients as a part of providing basic materials for reduction of radiation exposure to PET-CT test patients. In theory the measurement of external radiation dose rate of PET-CT test patients shows that the further the distance from the patient injected with radioactive pharmaceutical and a longer time elapsement from the injection leads to a smaller amount of radiation. Particularly, the amount of radiation marked the highest in the chest was at 4.17 minutes immediately after the intravenous injection and in the head after 77.47 minutes after urination in advance to the PET-CT test. As in the generalized information, it is desired to keep distance between the patient and caretakers or professionals to reduce the amount of radiation exposure from PET-CT test patients and to resume contact the patient after the time when the radiation has reduced. If contact is unavoidable, it is desired to keep at least 200cm from the patient. In addition, the amount of radiation reached the highest in the chest at first and then in the head from 77 minutes after injection. Accordingly, it would be helpful in achieving the optimization if contact is made based on the patient's physical characteristics. This study is significant as it measures changes in radiation the dose rate by; distance from the PET-CT test patient, time elapsed, and specific parts of body. Further studies based on the findings in this paper are required to analyze changes in radiation dose rate in accordance with individual characteristics unique to PET-CT patients and to utilize the results to reduce the amount of radiation patient, caretakers and professions are exposed.

• Journal of Radiation Protection and Research
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• v.11 no.1
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• pp.57-64
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• 1986

This paper described an improved technical method required for proper evaluation of personnel exposures by means of the photographic dosimeter developed by KAERI in lower gamma or X-ray energy regions, with which response of the dosimeter varies significantly. With calibration of the dosimeter in the energy range from 30 to 300 keV, the beam spectrum was carefully selected and specified it adequately. The absorber combinations and absorber thickness used to obtain the specified X-ray spectra from a constant potential X-ray machine were determined theoretically and also experimentally. A correlation between the density and exposure for the four separate energies, such as $49\;keV_{eff},\;154\;keV_{eff}\;250\;keV_{eff}\;and\;662\;keV$, is experimentally determined. As a result, it can be directly evaluated the exposure from the measured response of dosimeter.

• The Journal of the Korea Contents Association
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• v.12 no.3
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• pp.244-250
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• 2012

The results of analyzing the difference between performances of individual dosimeters on this research subjecting the PLD and TLD, which are the official personal dosimeters, through dosimetry are as follows. After scanning the integral dose using an automatic scanner, the values of two devices that went through dose adjustment process had a statistical difference in TLD and PLD measurements under each filming conditions which were 70kVp, 200mA, 0.012sec and 42kVp, 100mA, and 0.012sec (p<0.001 and p<0.001 respectively). As for the difference of measurement value between DAP and the two particles under 70kVp, 200mA, 0.012sec filming condition, TLD had a value lower than DAP average value by $44.2mGy{\cdot}cm^2$ and PLD had a value of $246.8mGy{\cdot}cm^2$ which was lower than DAP average value by $15.5mGy{\cdot}cm^2$, while under 42kVp, 100mA, 0.012sec filming condition, TLD had a value lower than DAP average value by $17.9mGy{\cdot}cm^2$ and PLD had a value of $82.6mGy{\cdot}cm^2$ which was lower than DAP average value by 7.6$mGy{\cdot}cm^2$. Also, compared to PLD, each of 10 devices measured dose value in TLD had a larger deviation between the particles, and for a reproducibility test which repeatedly measured one particle, PLD had ${\pm}1%$ which was lower than TLD's ${\pm}2%$. As such, PLD had a superior performance result in dose measurement capacities aspect compared to TLD, and therefore we could verify that PLD is more appropriate and advantageous in managing radiation-related task performing worker's personal radiation exposure management in the diagnostic radiation field.

• Journal of Radiation Protection and Research
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• v.40 no.4
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• pp.216-222
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• 2015

Facilities processing raw materials containing naturally occurring radioactive materials (NORM) may give rise to enhanced radiation dose to workers due to chronic inhalation of airborne particulates. Internal radiation dose due to particulate inhalation varies depending on particulate properties, including size, shape, density, and absorption type. The objective of the present study was to assess inhalation dose sensitivity to physicochemical properties of airborne particulates. Committed effective doses to workers resulting from inhalation of airborne particulates were calculated based on International Commission on Radiological Protection 66 human respiratory tract model. Inhalation dose generally increased with decreasing particulate size. Committed effective doses due to inhalation of $0.01{\mu}m$ sized particulates were higher than doses due to $100{\mu}m$ sized particulates by factors of about 100 and 50 for $^{238}U$ and $^{230}Th$, respectively. Inhalation dose increased with decreasing shape factor. Shape factors of 1 and 2 resulted in dose difference by about 18 %. Inhalation dose increased with particulate mass density. Particulate mass densities of $11g{\cdot}cm^{-3}$ and $0.7g{\cdot}cm^{-3}$ resulted in dose difference by about 60 %. For $^{238}U$, inhalation doses were higher for absorption type of S, M, and F in that sequence. Committed effective dose for absorption type S of $^{238}U$ was about 9 times higher than dose for absorption F. For $^{230}Th$, inhalation doses were higher for absorption type of F, M, and S in that sequence. Committed effective dose for absorption type F of $^{230}Th$ was about 16 times higher than dose for absorption S. Consequently, use of default values for particulate properties without consideration of site specific physiochemical properties may potentially skew radiation dose estimates to unrealistic values up to 1-2 orders of magnitude. For this reason, it is highly recommended to consider site specific working materials and conditions and use the site specific particulate properties to accurately access radiation dose to workers at NORM processing facilities.