• Journal of Radiation Protection and Research
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• v.38 no.3
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• pp.157-165
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• 2013

To determine the factors affecting the external radiation dose rates of patients undergoing PET-MRI examinations and to assess the trends of these differences, we measured the changes in the dose rates of $^{18}F$-FDG during a set period of time for each body region. Consistent with theoretical predictions, the dose rate decreased over time in patients undergoing PET-MRI examinations. Furthermore, immediately after the $^{18}F$-FDG injection, the dose rate in the chest region was the highest, followed by the abdominal region, the head region, and the foot region. The dose rate decreased drastically as time passed, by 2.47-fold, from $339.23{\pm}74.70mSv\;h^{-1}$ ($6.73{\pm}5.79$ min) at the time point immediately after the $^{18}F$-FDG injection to $102.71{\pm}26.17mSv\;h^{-1}$ ($136.11{\pm}25.64$ min) after the examination. In the foot region, there were no significant changes over time, from $32.05{\pm}20.23mSv\;h^{-1}$ ($6.73{\pm}5.79$ min) at the time point immediately after the $^{18}F$-FDG injection, to $23.89{\pm}9.14mSv\;h^{-1}$ ($136.11{\pm}25.64$ min) after the examination. The dose rate is dependent on the individual characteristics of the patient, and differed depending on the body region and time point. However, the dose rates were higher in patients who had a lower body weight, shorter stature, fewer urinations, lower fluid intake, and history of diabetes mellitus. To decrease radiation exposure, it is difficult or impossible to change factors inherent to the patient, such as sex, age, height, body weight, obesity, and history of diabetes mellitus. However, factors which can be changed, such as the $^{18}F$-FDG dose, fasting time, fluid intake, number of urinations, and contrast agent dose can be controlled to minimize the external radiation exposure of the patient.

• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.231-236
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• 2012

This study derived measures to reduce exposure doses by identifying factors which affect the external radiation dose rate of patients treated with radiopharmaceuticals for PET-CT tests. The external radiation dose rates were measured on three parts of head, thorax and abdomen at a distance of 50cm from the surface of 60 PET-CT patients. It showed there are changes in factors affecting the external radiation dose rate over time after the administration of F-18 FDG. The external radiation dose rate was lower in the patients with more water intake than those with less water intake before the injection of radiopharmaceuticals at all three points: right after the injection of radiopharmaceuticals (average 4.17 mins), after the pre-PEET-CT urination step (average 77.47 mins), and right after the PET-CT test (average 114.15 mins). The study also found there is a need to increase the amount of water intake before the injection of radiopharmaceuticals in order to maintain a low external radiation dose rate in patients. This strategy is only possible under the assumption that the quality of the video has not changed after conducting this study on the relations between the image and quality. This study also found a need to use radiopharmaceuticals with the minimum amount needed for each patient because F-FDG doses affects the external radiation dose rate at the point right after the injection of radiopharmaceuticals. Urination frequency was the most significant factor to affect the external radiation dose rates at the point right after the PET-CT test and the point after the pre-PET-CT urination step. There is a need to realize the strategy to increase the urination frequency of patients to maintain the external radiation dose rate low (average 77.47 mins) before and after the injection of radiopharmaceuticals. In addition, at this point, there is a need to take advantage of personal strategies because the external radiation dose rate is lower if the fasting time is shorter, the contrast medium is used, and the amount of water intake is increased after the administration of radiopharmaceuticals. Finally this study found the need to be able to generalize these findings through an in-depth research on the factors affecting the external radiation dose rate, which includes radiopharmaceutical dose, urination frequency, the amount of water intake, fasting time and the use of contrast medium.

• Journal of Radiation Protection and Research
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• v.41 no.1
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• pp.1-6
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• 2016

Background: It is very difficult to set a regulatory guidance or criteria for the protection of non-human species from the ionizing radiation, because there are no generally or internationally accepted methods for demonstrating the compliance with such criteria. It is needed that Korea develop the primary dose rate standards for the protection of both aquatic and terrestrial biota in the near future. Materials and Methods: The potential dose rates due to both external and internal radiation exposures to marine organisms such as plaice/flounder, gray mullet, and brown seaweed collected within territorial seas around the Korean Peninsula were estimated. Results and Discussion: The total dose rates to plaice/flounder, gray mullet and brown seaweed due to $^{40}K$, a primordial radionuclide in marine environment, were found to be 0.2%, 0.08% and 0.3% of approximately the values of the Derived Consideration Reference Levels (DCRLs, i.e. $1-10mGy{\cdot}d^{-1}$), respectively, as suggested by the International Commission on Radiological Protection (ICRP) publication 124. The total dose rates to marine fishes and brown seaweed due to anthropogenic radionuclides such as $^{90}Sr$, $^{137}Cs$ and $^{239+240}Pu$ were considered to be negligible compared to the total dose rate due to $^{40}K$. The external exposure to benthic fish due to all radionuclides was much higher than that of pelagic fish. Conclusion: From this study, it is recommended that the further study is required to develop a national regulatory guidance for the evaluation of doses to non-human species.

• Journal of radiological science and technology
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• v.46 no.2
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• pp.131-139
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• 2023

High-dose I-131 used for the treatment of thyroid cancer causes localized exposure among radiology technologists handling it. There is a delay between the calibration date and when the dose of I-131 is administered to a patient. Therefore, it is necessary to directly measure the radioactivity of the administered dose using a dose calibrator. In this study, we attempted to apply machine learning modeling to measured external dose rates from shielded I-131 in order to predict their radioactivity. External dose rates were measured at 1 m, 0.3 m, and 0.1 m distances from a shielded container with the I-131, with a total of 868 sets of measurements taken. For the modeling process, we utilized the hold-out method to partition the data with a 7:3 ratio (609 for the training set:259 for the test set). For the machine learning algorithms, we chose linear regression, decision tree, random forest and XGBoost. To evaluate the models, we calculated root mean square error (RMSE), mean square error (MSE), and mean absolute error (MAE) to evaluate accuracy and R² to evaluate explanatory power. Evaluation results are as follows. Linear regression (RMSE 268.15, MSE 71901.87, MAE 231.68, R² 0.92), decision tree (RMSE 108.89, MSE 11856.92, MAE 19.24, R² 0.99), random forest (RMSE 8.89, MSE 79.10, MAE 6.55, R² 0.99), XGBoost (RMSE 10.21, MSE 104.22, MAE 7.68, R² 0.99). The random forest model achieved the highest predictive ability. Improving the model's performance in the future is expected to contribute to lowering exposure among radiology technologists.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.378-383
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• 2016

Background: Radiation dose rates in PRIDE facility is evaluated quantitatively for assessing radiation safety of workers because of large amounts of depleted uranium being handled in PRIDE facility. Even if direct radiation from depleted uranium is very low and will not expose a worker to significant amounts of external radiation. Materials and Methods: ORIGEN-ARP code was used for calculating the neutron and gamma source term being generated from depleted uranium (DU), and the MCNP5 code was used for calculating the neutron and gamma fluxes and dose rates. Results and Discussion: The neutron and gamma fluxes and dose rates due to DU on spherical surface of 30 cm radius were calculated with the variation of DU mass and density. In this calculation, an imaginary case in which DU density is zero was added to check the self-shielding effect of DU. In this case, the DU sphere was modeled as a point. In case of DU mixed with molten salt of 50-250 g, the neutron and gamma fluxes were calculated respectively. It was found that the molten salt contents in DU had little effect on the neutron and the gamma fluxes. The neutron and the gamma fluxes, under the respective conditions of 1 and 5 kg mass of DU, and 5 and $19.1g{\cdot}cm^{-3}$ density of DU, were calculated with the molten salt (LiCl+KCl) of 50 g fixed, and compared with the source term. As the results, similar tendency was found in neutron and gamma fluxes with the variation of DU mass and density when compared with source spectra, except their magnitudes. Conclusion: In the case of the DU mass over 5 kg, the dose rate was shown to be higher than the environmental dose rate. From these results, it is concluded that if a worker would do an experiment with DU having over 5 kg of mass, the worker should be careful in order not to be exposed to the radiation.

• The Korean Journal of Nuclear Medicine Technology
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• v.22 no.1
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• pp.67-75
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• 2018

Purpose The purpose of this study is to investigate and analyze the external dose rates of $^{18}F-FDG$ Whole Body PET/CT patients by distance, and to identify the main factors that contribute to the reduction of radiation dose by checking the cumulative doses of nuclear medicine technologist(NMT). Materials and Methods After completion of the $^{18}F-FDG$ Whole Body PET/CT scan($75.4{\pm}3.3min$), the external dose rates of 106 patients were measured at a distance of 0, 10, 30, 50, and 100 cm from the chest. Gender, age, BMI(Body Mass Index), fasting time, diabetes mellitus, radiopharmaceutical injection information, creatine value were collected to analyze individual factors that could affect external dose rates from a patient's perspective. From the perspective of NMT, personal pocket dosimeters were worn on the chest to record accumulated dose of NMT who performed the injection task($T_1$, $T_2$ and $T_3$) and scan task($T_4$, $T_5$ and $T_6$). In addition, patient contact time with NMT was measured and analyzed. Results External dose rates from the patient for each distance were calculated as $246.9{\pm}37.6$, $129.9{\pm}16.7$, $61.2{\pm}9.1$, $34.4{\pm}5.9$, and $13.1{\pm}2.4{\mu}Sv/hr$ respectively. On the patient's aspect, there was a significant difference in the proximity of gender, BMI, Injection dose and creatine value, but the difference decreased as the distance increased. In case of dialysis patient, external dose rates for each distance were exceptionally higher than other patients. On the NMT aspect, the doses received from patients were 0.70, 1.09, $0.55{\mu}Sv/person$ for performing the injection task($T_1$, $T_2$, and $T_3$), and were 1.25, 0.82, $1.23{\mu}Sv/person$ for performing the scan task($T_4$, $T_5$, $T_6$). Conclusion we found that maintaining proper distance with patient and reducing contact time with patient had a significant effect on accumulated doses. Considering those points, efforts such as sufficient water intake and encourage of urination, maintaining the proper distance between the NMT and the patient(at least 100 cm), and reducing the contact time should be done for reducing dose rates not only patient but also NMT.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.547-556
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• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.181-190
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• 2012

Naturally occurring radioactive materials (NORM) in building materials are main sources of external radiation exposure to the general public. The objective of this study was to assess external radiation dose in Korean dwellings due to NORM in concrete walls. Reference room model for dose assessment was made by analyzing room structure and housing scale of Korean dwellings. In addition, dose assessments were made for varying room sizes. Absorbed doses to air and effective dose rates were calculated using radiation transport code MCNPX. Assuming a reference room of $3{\times}4{\times}2.8m^3$, absorbed dose rates in air were 0.80, 0.97, 0.08 nGy $h^{-1}$ per Bq $kg^{-1}$ for uranium series, thorium series, and $^{40}K$, respectively. Effective dose rates were 0.57, 0.69, 0.058 nSv $h^{-1}$ per Bq $kg^{-1}$, respectively. Radiation dose resulting from concrete of ceiling and floor increased with room area while radiation dose from concrete of walls decreased with room area. Therefore, total radiation doses were almost the same for the varying room area from 5 to $30m^2$. Effective dose in Korean dwellings was calculated based on measurement data of NORM concentration in concrete and occupancy fraction of Korean population by location. Annual effective dose was 0.59 mSv assuming that indoor occupancy fraction was 0.89 and concentrations of uranium series, thorium series and $^{40}K$ were 26, 39, 596 Bq $kg^{-1}$, respectively. Finally, annual effective dose in Korean dwellings can be calculated by the following equation: Effective dose=indoor occupancy fraction${\times}8760\;h\;y^{-1}{\times}(0.57C_U+0.69C_{Th}+0.058C_K)$.

• Journal of radiological science and technology
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• v.38 no.1
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• pp.23-30
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• 2015

Positron emission tomography scan has been growing diagnostic equipment in the development of medical imaging system. Compare to 99mTc emitting 140 keV, Positron emission radionuclide emits 511 keV gamma rays. Because of this high energy, it needs to reduce radioactive emitting from patients for radio technologist. We searched the external dose rates by changing distance from patients and measure the external dose rates when we used shielder investigate change external dose rates. In this study, the external dose distribution were analyzed in order to help managing radiation protection of radio technologists. Ten patients were searched (mean age: $47.7{\pm}6.6$, mean height: $165.5{\pm}3.8cm$, mean weight: $65.9{\pm}1.4kg$). Radiation was measured on the location of head, chest, abdomen, knees and toes at the distance of 10, 50, 100, 150, and 200 cm, respectively. Then, all the procedure was given with a portable radiation shielding on the location of head, chest, and abdomen at the distance of 100, 150, and 200 cm and transmittance was calculated. In 10 cm, head ($105.40{\mu}Sv/h$) was the highest and foot($15.85{\mu}Sv/h$) was the lowest. In 200 cm, head, chest, and abdomen showed similar. On head, the measured dose rates were $9.56{\mu}Sv/h$, $5.23{\mu}Sv/h$, and $3.40{\mu}Sv/h$ in 100, 150, and 200 cm, respectively. When using shielder, it shows $2.24{\mu}Sv/h$, $1.67{\mu}Sv/h$, and $1.27{\mu}Sv/h$ in 100, 150, and 200 cm on head. On chest, the measured dose rates were $8.54{\mu}Sv/h$, $4.90{\mu}Sv/h$, $3.44{\mu}Sv/h$ in 100, 150, and 200 cm, respectively. When using shielder, it shows $2.27{\mu}Sv/h$, $1.34{\mu}Sv/h$, and $1.13{\mu}Sv/h$ in 100, 150, and 200 cm on chest. On abdomen, the measured dose rates were $9.83{\mu}Sv/h$, $5.15{\mu}Sv/h$, and $3.18{\mu}Sv/h$ in 100, 150, and 200 cm, respectively. When using shielder, it shows $2.60{\mu}Sv/h$, $1.75{\mu}Sv/h$, and $1.23{\mu}Sv/h$ in 100, 150, and 200 cm on abdomen. Transmittance was increased as the distance was expanded. As the distance was further, the radiation dose were reduced. When using shielder, the dose were reduced as one-forth of without shielder. The Radio technologists are exposed of radioactivity and there were limitations on reducing the distance with Therefore, the proper shielding will be able to decrease radiation dose to the technologists.

• The Korean Journal of Nuclear Medicine Technology
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• v.14 no.2
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• pp.159-165
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• 2010

Purpose: Positron emission tomography scan has been growing diagnostic equipment in the development of medical imaging system. Compare to $^{99m}Tc$ emitting 140 keV, Positron emission radionuclide emits 511 keV gamma rays. Because of this high energy, it needs to reduce radioactive emitting from patients for radiotechnologist. We searched the external dose rates by changing distance from patients and measure the external dose rates when we used shielder investigate change external dose rates. In this study, the external dose distribution were analyzed in order to help managing radiation protection of radiotechnologists. Materials and Methods: Ten patients were searched (mean age: $47.7{\pm}6.6$, mean height: $165.5{\pm}3.8$ cm and mean weight: $65.9{\pm}1.4$ kg). Radiation were measured on the location of head, chest, abdomen, knees and toes at the distance of 10, 50, 100, 150 and 200 cm. Then, all the procedure was given with a portable radiation shielding on the location of head, chest and abdomen at the distance of 100, 150 and 200 cm and transmittance was calculated. Results: In 10 cm, head (105.40 ${\mu}Sv/h$) was the highest and foot (15.85 ${\mu}Sv/h$) was the lowest. In 200 cm, head, chest and abdomen showed similar. On head, the measured dose rates were 9.56 ${\mu}Sv/h$, 5.23 ${\mu}Sv/h$, and 3.40 ${\mu}Sv/h$ in 100, 150 and 200 cm respectively. When using shielder, it shows 2.24 ${\mu}Sv/h$, 1.67 ${\mu}Sv/h$, and 1.27 ${\mu}Sv/h$ in 100, 150 and 200 cm on head. On chest, the measured dose rates were 8.54 ${\mu}Sv/h$, 4.90 ${\mu}Sv/h$, 3.44 ${\mu}Sv/h$ in 100, 150 and 200 cm, respectively. When using shielder, it shows 2.27 ${\mu}Sv/h$, 1.34 ${\mu}Sv/h$, and 1.13 ${\mu}Sv/h$ in 100, 150 and 200 cm on chest. On abdomen, the measured dose rates were 9.83 ${\mu}Sv/h$, 5.15 ${\mu}Sv/h$ and 3.18 ${\mu}Sv/h$ in 100, 150 and 200cm respectively. When using shielder, it shows 2.60 ${\mu}Sv/h$, 1.75 ${\mu}Sv/h$ and 1.23 ${\mu}Sv/h$ in 100, 150 and 200 cm on abdomen. Transmittance was increased as the distance was expanded. Conclusion: As the distance was further, the radiation dose were reduced. When using shielder, the dose were reduced as one-forth of without shielder. The Radio technologists are exposed of radioactivity and there were limitations on reducing the distance with Therefore, the proper shielding will be able to decrease radiation dose to the radiotechnologists.