• The Korean Journal of Nuclear Medicine Technology
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• v.12 no.1
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• pp.91-98
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• 2008

Purpose: Radio-isotopes (RI) use has been steadily developing due to industrial and technical development in the modern medical society. Particularly, popularization of domestic cyclotrons dramatically enable hospitals to produce and use diagnostic radio-isotopes. Generally, only specific facilities such as hospitals, research institutes, nuclear power plants and universities can use radio-isotopes, they are also responsible for ventilation system. The strength of radioactivity in the air is strongly regulated and controlled by korea atomic energy law in Korea Institue of Nuclear Safety (KINS), so that air radioactivity exposure can lead to environmental pollution surrounding places. In this study, we'd like to find out the investigation and the present condition of the controlled ventilation system in domestic hospitals by an emission standard from KINS, and try to reach an agreement about how to use the ventilation system. Result: Definition of filters, features and structures of pre-filters, hepa-filters, charcol filters, filter exchange procedures and precautions are explained. RI deflation concentration and filter exchange cycle have been presented as a standard prescribed in the rules of KINS. The Radiation Control Management System (RCMS) introduced by Seoul National University Bundang Hospital linking to digital pressure gauge with computer controller in another medical facilities were described in details. Conclusions: The system of medical facilities using RI has been remarkably developing in 21 century. Especially, radiation safety control system has also been grown rapidly into the subdivision, specialization, advanced technology along with international technical improvement. However, As far as current RI ventilation system is concerned, it has nothing better than doing in the past. Preferentially, to reinforce this, more sophisticated system with strict periodic filter exchange and exhaust air control guidance should be introduced by applying brilliant domestic information technology for RCMS and digital gauge method. From personal point of view as a radiation safety manager, I have provide with present problems and improvements. Futhermore, more improved guidance should be conducted.

• The Korean Journal of Nuclear Medicine Technology
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• v.18 no.2
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• pp.22-27
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• 2014

Purpose There is the recent PET/CT scan in tendency that use low dose to reduce patient's exposure along with development of equipments. We diminished $^{18}F$-FDG dose of patient to reduce patient's exposure after setting up GE Discovery 690 PET/CT scanner (GE Healthcare, Milwaukee, USA) establishment at this hospital in 2011. Accordingly, We evaluate acquisition time per proper bed by change of infusion dose to maintain quality of image of PET/CT scanner. Materials and Methods We inserted Air, Teflon, hot cylinder in NEMA NU2-1994 phantom and maintained radioactivity concentration based on the ratio 4:1 of hot cylinder and back ground activity and increased hot cylinder's concentration to 3, 4.3, 5.5, 6.7 MBq/kg, after acquisition image as increase acquisition time per bed to 30 seconds, 1 minute, 1 minute 30 seconds, 2 minute, 2 minutes 30 seconds, 3 minutes, 3 minutes 30 seconds, 4 minutes, 4 minutes 30 seconds, 5 minutes, 5 minutes 30 seconds, 10 minutes, 20 minutes, and 30 minutes, ROI was set up on hot cylinder and back radioactivity region. We computated standard deviation of Signal to Noise Ratio (SNR) and BKG (Background), compared with hot cylinder's concentration and change by acquisition time per bed, after measured Standard Uptake Value maximum ($SUV_{max}$). Also, we compared each standard deviation of $SUV_{max}$, SNR, BKG following in change of inspection waiting time (15minutes and 1 hour) by using 4.3 MBq phantom. Results The radioactive concentration per unit mass was increased to 3, 4.3, 5.5, 6.7 MBqs. And when we increased time/bed of each concentration from 1 minute 30 seconds to 30 minutes, we found that the $SUV_{max}$ of hot cylinder acquisition time per bed changed seriously according to each radioactive concentration in up to 18.3 to at least 7.3 from 30 seconds to 2 minutes. On the other side, that displayed changelessly at least 5.6 in up to 8 from 2 minutes 30 seconds to 30 minutes. SNR by radioactive change per unit mass was fixed to up to 0.49 in at least 0.41 in 3 MBqs and accroding as acquisition time per bed increased, rose to up to 0.59, 0.54 in each at least 0.23, 0.39 in 4.3 MBqs and in 5.5 MBqs. It was high to up to 0.59 from 30 seconds in radioactivity concentration 6.7 MBqs, but kept fixed from 0.43 to 0.53. Standard deviation of BKG (Background) was low from 0.38 to 0.06 in 3 MBqs and from 2 minutes 30 seconds after, low from 0.38 to 0 in 4.3 MBqs and 5.5 MBqs from 1 minute 30 seconds after, low from 0.33 to 0.05 in 6.7 MBqs at all section from 30 seconds to 30 minutes. In result that was changed the inspection waiting time to 15 minutes and 1 hour by 4.3 MBq phantoms, $SUV_{max}$ represented each other fixed values from 2 minutes 30 seconds of acquisition time per bed and SNR shown similar values from 1 minute 30 seconds. Conclusion As shown in the above, when we increased radioactive concentration per unit mass by 3, 4.3, 5.5, 6.7 MBqs, the values of $SUV_{max}$ and SNR was kept changelessly each other more than 2 minutes 30 seconds of acquisition time per bed. In the same way, in the change of inspection waiting time (15 minutes and 1 hour), we could find that the values of $SUV_{max}$ and SNR was kept changelessly each other more than 2 minutes 30 seconds of acquisition time per bed. In the result of this NEMA NU2-1994 phantom experiment, we found that the minimum acquisition time per bed was 2 minutes 30 seconds for evaluating values of fixed $SUV_{max}$ and SNR even in change of inserting radioactive concentration. However, this acquisition time can be different according to features and qualities of equipment.

• The Korean Journal of Nuclear Medicine Technology
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• v.16 no.2
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• pp.12-17
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• 2012

Purpose : The patient's clothes and sheet after radioiodine therapy must be disposed of by related regulation. That must be disposed of as radioactive wastes, but that is reusing after radioactivity decay by keeping for the certain period of time. In general, The minimum storage period calculate by standard of take radioactive substance out of radiation controlled area based on measured surface contamination level. But the measurements of surface contamination level are able to differ by measurement method. In this paper, I wish to calculate the minimum storage period of patient's clothes and sheet after radioiodine therapy by measure nuclide concentration offered by the regulation on self-disposal of radioactive wastes. Materials and Methods : The whole area of patient's clothes and sheet measured 31 patients(male:9 patients, female:22 patients), who had radioiodine therapy(3.7 GBq:13 patients, 5.55 GBq:16 patients, 7.4 GBq:2 patients) from july 2011 to march 2012. The minimum storage period is calculated by the regulation on self-disposal of radioactive waste(100 Bq/g) and standard of take radioactive substance out of radiation controlled area(4 kBq/m2) Results : The minimum storage period of pillow sheet, upper uniform, lower uniform by standard of take radioactive substance out of radiation controlled area were each 4.6 days, 63days, 78 days. The minimum storage period of pillow sheet, upper uniform, lower uniform by the regulation on self-disposal of radioactive waste were each 18.1 days, 43 days, 62 days. Conclusion : We can verify that patient's clothes and sheet after radioiodine therapy exists a great deal of radioactive contamination. The minimum storage period calculation of patient's clothes and sheet is better suited to applying nuclide concentration offered by the regulation on self-disposal of radioactive waste. I recommend, To keep for at least 2 months of the patient's clothes and sheet contaminated radioactivity, for prevent contamination and unnecessary radiation exposure.

• Journal of Pharmaceutical Investigation
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• v.33 no.2
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• pp.145-149
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• 2003

Manufacturing facilities of the pharmaceuticals must meet certain level of the cleanness required so that foreign substances such as dust, moisture, heat, microorganism, or virus do not contaminate the product. In case of radiopharmaceuticals for medical treatment and diagnosis, not only should the operators and environment be protected from radiation but also need to be isolated from the foreign contaminant. Therefore, manufacturing facilities for radiopharmaceuticals must satisfy the design standards of both hot cell and clean room which are specified by GMP. However, standards of maintaining negative pressure for preventing spread of radioactive contaminant in isolated facilities conflict with the standards of maintaining positive pressure for keeping cleanness. To solve this problem, air pressure of hot cell was designed lower than in the adjacent area to meet standards of the radiation safety. To keep higher cleanness in certain part of the hot cell for filling, minimal relative positive pressure allows. In order to effectively maintain the cleanness that is required for production of Tc-99m generator, which takes 70% of whole demand of radiopharmaceuticals, the rooms placed in each side of production room are used as a buffer area and three lead hot cells are installed in production room. In this research, we established the appropriate engineered design concept for Tc-99m generator manufacturing facility, which satisfies both GMP cleanness standard for preventing particles, bacteria, other contaminants and the regulations of radiation safety for supervising and controlling the amount of radiation exposure and exhausted radioactivity. And the concept of multi-barrier buffer zones is introduced to apply negative air pressure for hot cell with first priority and to continue relative positive air pressure for clean room.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.06a
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• pp.84-85
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• 2009

New approaches for detecting, preventing and remedying environmental damage are important for protection of the environment. Procedures must be developed and implemented to reduce the amount of waste produced in chemical processes, to detect the presence and/or concentration of contaminants and decontaminate fouled environments. Contamination can be classified into three general types: airborne, surface and structural. The most dangerous type is airborne contamination, because of the opportunity for inhalation and ingestion. The second most dangerous type is surface contamination. Surface contamination can be transferred to workers by casual contact and if disturbed can easily be made airborne. The decontamination of the surface in the nuclear facilities has been widely studied with particular emphasis on small and large surfaces. The amount of wastes being produced during decommissioning of nuclear facilities is much higher than the total wastes cumulated during operation. And, the process of decommissioning has a strong possibility of personal's exposure and emission to environment of the radioactive contaminants, requiring through monitoring and estimation of radiation and radioactivity. So, it is important to monitor the radioactive contamination level of the nuclear facilities for the determination of the decontamination method, the establishment of the decommissioning planning, and the worker's safety. But it is very difficult to measure the surface contamination of the floor and wall in the highly contaminated facilities. In this study, the poly(styrene-ethyl acrylate) [poly(St-EA)] core-shell composite polymer for measurement of the radioactive contamination was synthesized by the method of emulsion polymerization. The morphology of the poly(St-EA) composite emulsion particle was core-shell structure, with polystyrene (PS)as the core and poly(ethyl acrylate) (PEA) as the shell. Core-shell polymers of styrene (St)/ethyl acrylate (EA) pair were prepared by sequential emulsion polymerization in the presence of sodium dodecyl sulfate (SOS) as an emulsifier using ammonium persulfate (APS) as an initiator. The polymer was made by impregnating organic scintillators, 2,5-diphenyloxazole (PPO) and 1,4-bis[5-phenyl-2-oxazol]benzene (POPOP). Related tests and analysis confirmed the success in synthesis of composite polymer. The products are characterized by IT-IR spectroscopy, TGA that were used, respectively, to show the structure, the thermal stability of the prepared polymer. Two-phase particles with a core-shell structure were obtained in experiments where the estimated glass transition temperature and the morphologies of emulsion particles. Radiation pollution level the detection about under using examined the beta rays. The morphology of the poly(St-EA) composite polymer synthesized by the method of emulsion polymerization was a core-shell structure, as shown in Fig. 1. Core-shell materials consist of a core structural domain covered by a shell domain. Clearly, the entire surface of PS core was covered by PEA. The inner region was a PS core and the outer region was a PEA shell. The particle size distribution showed similar in the range 350-360 nm.

• Progress in Medical Physics
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• v.27 no.3
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• pp.156-161
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• 2016

When air discharged from a radioisotope production facility is contaminated with radiation, the public may be exposed to radiation. The objective of this study is to manage such radiation exposure. We measured the airborne radioactivity concentration at a 30 MeV cyclotron radioisotope production facility to assess whether the exhaust gas was contaminated. Additionally, we investigted the radioactive contamination of the air filter for efficient air purification and radiation safety control. To measure the airborne radiation concentration, specimens were collected weekly for 4 h after the beginning of the radioisotope production. Regarding the air purifier, five specimens were collected at different positions of each filter-pre-filter, high-efficiency particulate air filter, and charcoal filter-installed in the cyclotron production room. The concentrations of F-18, I-123, I-131, and Tl-201 generated in the radioiodine production room were $13.5Bq/m^3$, $27.0Bq/m^3$, $0.10Bq/m^3$, and $11.5Bq/m^3$, respectively; the concentrations of F-18, I-123, and I-131 produced in the radioisotope production room were $0.05Bq/m^3$, $16.1Bq/m^3$, and $0.45Bq/m^3$, correspondingly; and those of F-18, I-123, I-131, and Tl-201 generated in the accelerator room were $2.07Bq/m^3$, $53.0Bq/m^3$, $0.37Bq/m^3$, and $0.15Bq/m^3$, respectively. The maximum radiation concentration of I-123 generated in the radioiodine production room was 1,820 Bq/g, which can be disposed after 2 days. The maximum radiation concentration of Tl-202 generated in the radioisotope production room was 205 Bq/g, and this isotope must be stored for 53 days. The I-123 generated in the radioiodine production room had a maximum concentration of 1,530 Bq/g and must be stored for 2 days. The maximum radiation concentration of Na-22 generated in the radioisotope production room was 0.18 Bq/g and this isotope must be disposed after 827 days. To manage the exhaust, the efficiency of air purification must be enhanced by selecting an air purifier with a long life and determining the appropriate replacement time by examining the differential pressure through systematic measurements of the airborne radiation contamination level.

• The Korean Journal of Pesticide Science
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• v.8 no.1
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• pp.38-45
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• 2004

To elucidate the photolysis characteristics of the insecticide imidacloprid in the environment, $[^{14}C]$imidacloprid was treated into water and paddy soil-water system. In water system, the amount of $^{14}C$-radioactivity distributed in aqueous phase was rapidly increased up to 80% of total $^{14}C$ in water during 7 days of exposure to sunlight. Also, the amounts of imidacloprid in water at day 0 and 3 days after treatment were 1.2461 and 0.8594 mg/kg, respectively, not being detected 7 days after treatment, indicating rapid degradation of imidacloprid in water by sunlight. One photodegradation product, imidacloprid urea, in which the $N-NO_2$ moiety of imidacloprid was replaced by oxygen, was detected from water in water and water-paddy systems. The amount of the metabolite detected from water in water system was 0.0112 mg/kg 1 day after treatment and reached the top concentration of 0.0391 mg/kg 7 days after treatment. In case of water-paddy system, its amount was 0.0117 mg/kg 1 day after treatment and reached the highest concentration of 0.0259 mg/kg 3 days after treatment. Rapid transformation of imidacloprid into polar compounds continued until 7 days after treatment, considering that 80% of $^{14}C$ in water distributed in aqueous phase 7 days after treatment, amount of imidacloprid was 1.6538 mg/kg at day 0 and 0.8785 mg/kg 1 day after treatment, not being detected after 15 days, indicating rapid degradation of imidacloprid in water-paddy soil system by sunlight. The direct degradation of imidacloprid to imidacloprid urea would be a major photodegradation pathway in water and water-paddy soil systems.

• Journal of Radiation Protection and Research
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• v.33 no.1
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• pp.21-26
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• 2008

When the calibration on Liquid Scintillation Counter using the Solid $^3H$ Standard Source of 200,000DPM is executed, the uncertainty due to activity and geometry difference, exists. Therefore, this paper intends to evaluate environmental samples comparatively accurately as decreasing this uncertainty existing in the process of calibration. For this, measurements on samples manufactured by $^3H$ Standard Source and sensitivity study were performed. Also, this paper verified calibration results using Radioactivity-Error-Analysis Method, and evaluated quantitatively the effect by geometry and activity difference based on verification result. According to the result of sensitivity study, in case of using the exposure time of 75 sec and Repeat method, the measuring accuracy and precision of about $1{\sim}3%$ were increased in comparison with the existing method. By analysis result, the effect by activity difference did not appear, and a plastic cell existing into Teflon vial made a role as reflector. The less the effect of plastic cells are decreased, the more activity is high, and the effect of those can be neglected at the activity of 200,000 DPM.

• The Korean Journal of Nuclear Medicine Technology
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• v.20 no.1
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• pp.59-65
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• 2016

Purpose Dynamic kidney scan is a typical imaging technique that visualizes kidney function. Reproducibility of dynamic kidney scans has been evaluated by comparing low-dose kidney scans with low-dose radiopharmaceutical and standard dynamic kidney scan. With this comparative study, if reproducibility is superb, the dynamic kidney scan method with reduced radioactivity to patients is to be utilized and radiation exposure to patient is to be reduced. Materials and Methods For gamma camera, Orbiter, SymbiaE (Siemens, Germany) was used. Among patients who had used 370 Mbq (10 mCi) from January of 2013 to February 2014 and other patients who had used 185 Mbq (5 mCi) from March of 2014 to July of 2015 with identical condition, 21 subjects using DTPA and 20 subjects using $MAG_3$, 41 subjects in total, had been selected as subjects for data. From renogram of the result image, frame of the peak point was selected. Then, region of interest of kidney and background had been selected and Kidney to Background Ratio has been calculated for comparison. Results In tests using DTPA, kidney to background ratio when using 370 Mbq was $5.67{\pm}0.8$ at average while it was $5.62{\pm}0.87$ when using 185 Mbq, which didn't show much difference. Also in the tests using $MAG_3$, kidney to background ratio when using 370 Mbq was $14.95{\pm}2.58$ at average and $14.56{\pm}2.02$ in 185 Mbq, which neither showed much difference. In paired sample t-test, p-value was 0.566 in DTPA and 0.363 in $MAG_3$, which confirmed that there was no difference between the groups. Conclusion In identical patients, when dose was decreased from 370 Mbq to 185 Mbq, reproducibility of dynamic kidney scan was proven to be excellent. Low-dose Dynamic kidney scan can achieve results with fine reproducibility without improvement in performance of gamma camera and is expected to reduce radiation exposure to patient.

• The Korean Journal of Nuclear Medicine Technology
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• v.22 no.2
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• pp.79-83
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• 2018

Purpose $^{18}F-FDG$, which is commonly used in PET-CT examinations, is low in capacity and it is difficult to keep the amount of radioactivity busy when the specific activity is high, increasing the amount of space dose and radioactive contamination in the distribution room. Therefore, while evaluating the actual dose administered to patients during the manual dispense process, the medical institution intends to assess the usefulness of the auto dispenser by comparing the differences from the actual dose administered to the patient using the new automatic dispense. Materials and Methods From July 2016 to December 2016, 846 patients were manually administered by workers using $^{18}F-FDG$ and $^{18}F-FDG$ 906 patients were using auto dispenser from July 2017 to December 2017. Results Capacity administered to patients during the manual dispense averaged $35.41{\pm}27.79%$ compared to the recommended dose, and the auto dispenser process showed a small difference of $-2.15{\pm}3.99%$ compared to the recommended dose(p<0.05). Conclusion Working people did not have to touch radioactive medicines directly while they were busy in the auto dispenser, and because of the availability of other tasks far away, the time and distance to receive the exposure were also advantageous. It is believed that future use by many medical institutions will not only reduce the dose to patients but also help reduce the exposure dose to workers.