• Journal of Radiation Protection and Research
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• v.23 no.4
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• pp.211-218
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• 1998

The distributional effects of radioactive materials on external gamma exposure have been analyzed. An approximate method for estimating external gamma dose given from an arbitrary distribution of radioactive material has been developed. The minimum gamma exposure given from a point source is shown at 0.07 MeV if the source to receptor distance is shorter than 10 m. But if the receptor to point source distance is longer than 20 m, gamma exposure rate increases monotonously according to the average gamma energy. For the analysis of the effects of volume source, we estimated the gamma dose given from different size of hemisphere in which radioactive materials are distributed uniformly. When the radius of hemisphere is longer than 40 m, external gamma dose rate increases monotonously. The gamma dose rate given from the radioactive materials deposited on the ground shows the minimum value at 0.07 MeV in any case. The analysis shows that external gamma exposure is strongly dependent on the distribution of radioactive materials in the environment and gamma energy.

• The Korean Journal of Nuclear Medicine Technology
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• v.20 no.1
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• pp.32-36
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• 2016

Purpose In nuclear medicine examination, $^{131}I$ is widely used in nuclear medicine examination such as diagnosis, treatment, and others of thyroid cancer and other diseases. $^{131}I$ conducts examination and treatment through emission of ${\gamma}$ ray and ${\beta}^-$ ray. Since $^{131}I$ (364 keV) contains more energy compared to $^{99m}Tc$ (140 keV) although it displays high integrated rate and enables quick discharge through kidney, the objective of this study lies in comparing the difference in exposure dose of $^{131}I$ before and after wearing apron when handling $^{131}I$ with focus on 3 elements of external exposure protection that are distance, time, and shield in order to reduce the exposure to technicians in comparison with $^{99m}Tc$ during the handling and administration process. When wearing apron (in general, Pb 0.5 mm), $^{99m}Tc$ presents shield of over 90% but shielding effect of $^{131}I$ is relatively low as it is of high energy and there may be even more exposure due to influence of scattered ray (secondary) and bremsstrahlung in case of high dose. However, there is no special report or guideline for low dose (74 MBq) high energy thus quantitative analysis on exposure dose of technicians will be conducted based on apron wearing during the handling of $^{131}I$. Materials and Methods With patients who visited Department of Nuclear Medicine of our hospital for low dose $^{131}I$ administration for thyroid cancer and diagnosis for 7 months from Jun 2014 to Dec 2014 as its subject, total 6 pieces of TLD was attached to interior and exterior of apron placed on thyroid, chest, and testicle from preparation to administration. Then, radiation exposure dose from $^{131}I$ examination to administration was measured. Total procedure time was set as within 5 min per person including 3 min of explanation, 1 min of distribution, and 1 min of administration. In regards to TLD location selection, chest at which exposure dose is generally measured and thyroid and testicle with high sensitivity were selected. For preparation, 74 MBq of $^{131}I$ shall be distributed with the use of $2m{\ell}$ syringe and then it shall be distributed after making it into dose of $2m{\ell}$ though dilution with normal saline. When distributing $^{131}I$ and administering it to the patient, $100m{\ell}$ of water shall be put into a cup, distributed $^{131}I$ shall be diluted, and then oral administration to patients shall be conducted with the distance of 1m from the patient. The process of withdrawing $2m{\ell}$ syringe and cup used for oral administration was conducted while wearing apron and TLD. Apron and TLD were stored at storage room without influence of radiation exposure and the exposure dose was measured with request to Seoul Radiology Services. Results With the result of monthly accumulated exposure dose of TLD worn inside and outside of apron placed on thyroid, chest, and testicle during low dose $^{131}I$ examination during the research period divided by number of people, statistics processing was conducted with Wilcoxon Signed Rank Test using SPSS Version. 12.0K. As a result, it was revealed that there was no significant difference since all of thyroid (p = 0.345), chest (p = 0.686), and testicle (p = 0.715) were presented to be p > 0.05. Also, when converting the change in total exposure dose during research period into percentage, it was revealed to be -23.5%, -8.3%, and 19.0% for thyroid, chest, and testicle respectively. Conclusion As a result of conducting Wilcoxon Signed Rank Test, it was revealed that there is no statistically significant difference (p > 0.05). Also, in case of calculating shielding rate with accumulate exposure dose during 7 months, it was revealed that there is irregular change in exposure dose for inside and outside of apron. Although the degree of change seems to be high when it is expressed in percentage, it cannot be considered a big change since the unit of accumulated exposure dose is in decimal points. Therefore, regardless of wearing apron during high energy low dose $^{131}I$ administration, placing certain distance and terminating the administration as soon as possible would be of great assistance in reducing the exposure dose. Although this study restricted $^{131}I$ administration time to be within 5 min per person and distance for oral administration to be 1m, there was a shortcoming to acquire accurate result as there was insufficient number of N for statistics and it could be processed only through non-parametric method. Also, exposure dose per person during lose dose $^{131}I$ administration was measured with accumulated exposure dose using TLD rather than through direct-reading exposure dose thus more accurate result could be acquired when measurement is conducted using electronic dosimeter and pocket dosimeter.

• Journal of Radiation Protection and Research
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• v.45 no.2
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• pp.53-68
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• 2020

Background: International organizations such as the World Health Organization (WHO) and the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) reported public exposure doses due to radionuclides released in the Fukushima nuclear accident a few years after the event. However, the reported doses were generally overestimated due to conservative assumptions such as a longer stay in deliberate areas designated for evacuation than the actual stay. After these reports had been published, more realistic dose values were reported by Japanese scientists. Materials and Methods: The present paper reviews those reports, including the most recently published articles; and summarizes estimated effective doses (external and internal) and issues related to their estimation. Results and Discussion: External dose estimation can be categorized as taking two approaches-estimation from ambient dose rate and peoples' behavior patterns-and measurements using personal dosimeters. The former approach was useful for estimating external doses in an early stage after the accident. The first 4-month doses were less than 2 mSv for most (94%) study subjects. Later on, individual doses came to be monitored by personal dosimeter measurements. On the basis of these measurements, the estimated median annual external dose was reported to be < 1 mSv in 2011 for 22 municipalities of Fukushima Prefecture. Internal dose estimation also can be categorized as taking two approaches: estimation from whole-body counting and estimation from monitoring of environmental samples such as radioactivity concentrations in food and drinking water. According to results by the former approach, committed effective dose due to ¹³⁴Cs and ¹³⁷Cs could be less than 0.1 mSv for most residents including those from evacuated areas. Conclusion: Realistic doses estimated by Japanese scientists indicated that the doses reported by WHO and UNSCEAR were generally overestimated. Average values for the first-year effective doses for residents in two affected areas (Namie Town and Iitate Village) were not likely to reach 10 mSv, the lower end of the doses estimated by WHO.

• Journal of the Korean Society of Radiology
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• v.8 no.4
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• pp.155-161
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• 2014

Mammography is mostly performed by series projection on both breasts. While taking mammography, it is less than average glandular dose of 3 mGy which regulated. But mammography is exposed much more doses actually, due to additional or series projection. Also, it is difficult to recognize around organ dose except exposed breast. Using mathematical simulation of radiation exposure body in mammography, we studied around organ dose distribution by changing thickness(25, 30, $50{\mu}m$) of filter and relative absorption dose rate which set on basis of exposed breast. as a result, when setting of basis of exposed breast, dose of opposite breast is more affected approximately from 79.26 to 86.31%. when using $25{\mu}m$ of filter thickness than $30{\mu}m$, $50{\mu}m$ of filter thickness in Mo/Mo, W/Rh combination which used actually, absorbed dose rates for opposite breast and around organ were low.

• Journal of Radiation Protection and Research
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• v.47 no.1
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• pp.30-38
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• 2022

Background: Residential areas have some factors on the external exposure of residents, who usually spend a long time in these areas. Although various survey has been carried out by the government or the research institutions after the Fukushima Daiichi Nuclear Power Plant accident, the mechanism of radiocesium inventory in the terrestrial zone has not been cleared. To better evaluate the radiation environment, this study investigated the temporal changes in air dose rate and ¹³⁷Cs inventories (Bq/m²) in residential areas and agricultural fields. Materials and Methods: Air dose rate and ¹³⁷Cs inventories were investigated in residential areas located in an evacuation zone at 5-8 km from the Fukushima Daiichi Nuclear Power Plant. From December 2014 to September 2018, the air dose rate distribution was investigated through a walking survey (backpack survey), which was conducted by operators carrying a γ-ray detector on their backs. Additionally, from December 2014 to January 2021, the ¹³⁷Cs inventories on paved and permeable grounds were also measured using a portable γ-ray detector. Results and Discussion: In the areas where decontamination was not performed, the air dose rate decreased faster in residential areas than in agricultural fields. Moreover, the ¹³⁷Cs inventory on paved surfaces decreased with time owing to the horizontal wash-off, while the ¹³⁷Cs inventory on permeable surfaces decreased dramatically owing to the decontamination activities. Conclusion: These findings suggest that the horizontal wash-off of ¹³⁷Cs on paved surfaces facilitated the air dose rate decrease in residential areas to a greater extent compared with agricultural fields, in which the air dose rate decreased because of the vertical migration of ¹³⁷Cs. Results of this study can explain the faster environmental restoration in a residential environment reported by previous studies.

• Journal of radiological science and technology
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• v.43 no.6
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• pp.503-509
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• 2020

This study aims to develop a patient-specific radiation exposure dose prediction model based on anthropometric data that can be easily measurable during CT examination, and to be used as basic data for DRL setting and radiation dose management system in the future. In addition, among the machine learning algorithms, the most suitable model for predicting exposure doses is presented. The data used in this study were chest CT scan data, and a data set was constructed based on the data including the patient's anthropometric data. In the pre-processing and sample selection of the data, out of the total number of samples of 250 samples, only chest CT scans were performed without using a contrast agent, and 110 samples including height and weight variables were extracted. Of the 110 samples extracted, 66% was used as a training set, and the remaining 44% were used as a test set for verification. The exposure dose was predicted through random forest, linear regression analysis, and SVM algorithm using Orange version 3.26.0, an open software as a machine learning algorithm. Results Algorithm model prediction accuracy was R^2 0.840 for random forest, R^2 0.969 for linear regression analysis, and R^2 0.189 for SVM. As a result of verifying the prediction rate of the algorithm model, the random forest is the highest with R^2 0.986 of the random forest, R^2 0.973 of the linear regression analysis, and R^2 of 0.204 of the SVM, indicating that the model has the best predictive power.

• KSBB Journal
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• v.28 no.6
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• pp.415-422
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• 2013

Gamma irradiation (${\gamma}IR$) is widely used for radiotherapy as a treatment of cancer cells although it has a risk to damage normal cells. Inflammation is regarded as one of side effects of ${\gamma}IR$ while the effect of low dose of ${\gamma}IR$ on inflammation has not been researched well. Here, we investigated the inflammatory responses of low dose of ${\gamma}IR$ on murine spleen cells. It was evaluated if ${\gamma}IR$ affected the mitogen-induced lymphocyte proliferation, the regulation of various inflammatory cytokines (IFN-${\gamma}$, IL-2, IL-17, IL-4, IL-10), and the involvement of Ikaros and MAPK/NF-${\kappa}B$ medicated mechanism. Exposure of $^{137}Cs-{\gamma}IR$ below 2 Gy decreased the lymphocytes proliferative response to mitogens (LPS, ConA) except at the lowest dose, 0.05 Gy. IL-17, IL-2 and IL-4 mRNA increased at 0.5 and 2 Gy, but not altered at 0.05 Gy. IL-10, anti-inflammatory cytokine, increased only at 0.05 Gy. In regard to intracellular signaling, p-JNK, p-p38 and p-$I{\kappa}B{\alpha}$ were not changed, whereas the activation of ERK and Ikaros increased at the lowest dose. These results suggest that exposure of ${\gamma}IR$ less than 0.5 Gy (or below 0.05 Gy) has beneficial effects as a radiation hormesis on immune function.

• Journal of Radiation Protection and Research
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• v.34 no.2
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• pp.55-64
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• 2009

In a pressurized heavy water reactor (PHWR), radiation workers who have access to radiation controlled areas submit their urine samples to health physicists periodically; internal radiation exposure is evaluated by the monitoring of these urine samples. Internal radiation exposure at PHWRs accounts for approximately 20 $\sim$ 40% of total radiation exposure; most internal radiation exposure is attributed to tritium. Carbon-14 is not a dominant nuclide in the radiation exposure of workers, but it is one potential nuclide to be necessarily monitored. Carbon-14 is a low energy beta emitter and passes relatively easily into the body of workers by inhalation because its dominant chemical form is radioactive carbon dioxide ($^{14}CO_2$). Most inhaled carbon-14 is rapidly exhaled from the worker's body, but a small amount of carbon-14 remains inside the body and is excreted by urine. In this study, a method for dual analysis of tritium and carbon-14 in urine samples of workers at nuclear power plants is developed and a method for internal dose assessment using its excretion rate result is established. As a result of the developed dual analysis of tritium and carbon-14 in urine samples of radiation workers who entered the high radiation field area at a PHWR, it was found that internal exposure to carbon-14 is unlikely to occur. In addition, through the urine counting results of radiation workers who participated in the open process of steam generators, it was found that the likelihood of internal exposure to either tritium or carbon-14 is extremely low at pressurized water reactors (PWRs).

• The Korean Journal of Nuclear Medicine Technology
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• v.18 no.1
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• pp.110-114
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• 2014

Purpose: The advancement in PET/CT test devices has decreased the test time and popularized the test, and PET/CT tests have continuously increased. However, this increases the exposure dose of radiation workers, too. This study aims to measure the radiation shielding rate of $^{18}F-FDG$ with a strong energy and the shielding effect when worker wore an apron during the PET/CT test. Also, this study compared the shielding rate with $^{99m}TC$ to minimize the exposure dose of radiation workers. Materials and Methods: This study targeted 10 patients who visited in this hospital for the PET/CT test for 8 days from May 2nd to 10th 2013, and the $^{18}F-FDG$ distribution room, patient relaxing room (stand by room after $^{18}F-FDG$ injection) and PET/CT test room were chosen as measuring spots. Then, the changes in the dose rate were measured before and after the application of the APRON. For an accurate measurement, the distance from patients or sources was fixed at 1M. Also, the same method applied to $^{99m}TC's$ Source in order to compare the reduction in the dose by the Apron. Results: 1) When there was only L-block in the $^{18}F-FDG$ distribution room, the average dose rate was $0.32{\mu}Sv$, and in the case of L-blockK+ apron, it was $0.23{\mu}Sv$. The differences in the dose and dose rate between the two cases were respectively, $0.09{\mu}Sv$ and 26%. 2) When there was no apron in the relaxing room, the average dose rate was $33.1{\mu}Sv$, and when there was an apron, it was $22.3{\mu}Sv$. The differences in the dose and dose rate between them were respectively, $10.8{\mu}Sv$ and 33%. 3) When there was no APRON in the PET/CT room, the average dose rate was $6.9{\mu}Sv$, and there was an APRON, it was $5.5{\mu}Sv$. The differences in the dose and dose rate between them were respectively, $1.4{\mu}Sv$ and 25%. 4) When there was no apron, the average dose rate of $^{99m}TC$ was $23.7{\mu}Sv$, and when there was an apron, it was $5.5{\mu}Sv$. The differences in the dose and dose rate between them were respectively, $18.2{\mu}Sv$ and 77%. Conclusion: According to the result of the experiment, $^{99m}TC$ injected into patients showed an average shielding rate of 77%, and $^{18F}FDG$ showed a relatively low shielding rate of 27%. When comparing the sources only, $^{18F}FDG$ showed a shielding rate of 17%, and $^{99m}TC$'s was 77%. Though it had a lower shielding effect than $^{99m}TC$, $^{18}F-FDG$ also had a shielding effect on the apron. Therefore, it is considered that wearing an apron appropriate for high energy like $^{18}F-FDG$ would minimize the exposure dose of radiation workers.

• Progress in Medical Physics
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• v.32 no.4
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• pp.145-152
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• 2021

Purpose: During the treatments of cancer patients with a linear accelerator (LINAC) using photon beams with energies ≥8 MV, the components inside the LINAC head get activated through the interaction of photonuclear reaction (γ, n) and neutron capture (n, γ). We used spectroscopy and measured the dose rate for the LINAC in operation after the treatment ended. Methods: We performed spectroscopy and dose rate measurements for three units of LINACs with a portable high-purity Germanium (HPGe) detector and a survey meter. The spectra were obtained after the beams were turned off. Spectroscopy was conducted for 3,600 seconds, and the dose rate was measured three times. We identified the radionuclides for each LINAC. Results: According to gamma spectroscopy results, most of the nuclides were short-lived radionuclides with half-lives of 100 days, except for ⁶⁰Co, ⁶⁵Zn, and ¹⁸¹W nuclides. The dose rate for three LINACs obtained immediately in front of the crosshair was in the range of 0.113 to 0.129 µSv/h. The maximum and minimum dose rates measured on weekends were 0.097 µSv/h and 0.092 µSv/h, respectively. Compared with the differences in weekday data, there was no significant difference between the data measured on Saturday and Sunday. Conclusions: Most of the detected radionuclides had half-lives <100 days, and the dose rate decreased rapidly. For equipment that primarily used energies ≤10 MV, when the equipment was transferred after at least 10 minutes after shutting it down, it is expected that there will be little effect on the workers' exposure.