• Journal of the Korean Magnetics Society
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• v.19 no.2
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• pp.52-56
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• 2009

Interlayer coupling field, coercivity, magnetoresitance ratio, and magnetic sensitivity depending on the thickness of free CoFe layer for the CoFe/Cu/CoFe/IrMn multilayer are investigated. In case of CoFe layer of $30\;{\AA}$ thickness for the CoFe(t)/Cu($25\;{\AA}$)/CoFe($60\;{\AA}$)/IrMn($80\;{\AA}$) multilayer with ferromagnet/non-magnet/ferromagnet structure induced by IrMn layer, the lowest coercivity and the highest magnetic sensitivity, which is contained soft magnetic property, are observed. On the other side, in case of CoFe layer of $90\;{\AA}$ thickness, there are the highest coercivity and the lowest magnetic sensitivity. The fabricated CoFe($30\;{\AA}$)/Cu($25\;{\AA}$)/CoFe($60\;{\AA}$)]/IrMn($80\;{\AA}$) spin valve device with $2{\times}8{\mu}m^2$ patterning size are measured by two probe method, which is selected the sensing current as the longitudinal direction and the easy axis as the transversal direction. The measuring magntoresistance ratio and magnetic sensitivity of GMR-SV device having the soft magnetic property are 3.0% and 0.3%/Oe, respectively.

• Journal of Radiation Industry
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• v.17 no.2
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• pp.151-160
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• 2023

The International Atomic Energy Agency (IAEA) proposes 11 industries that handle Naturally Occurring Radioactive Material (NORM) that are considered to need management. A water treatment facility is one of the above industries that takes in groundwater and produces drinking water through a water treatment process. Groundwater can accumulate natural radionuclides such as uranium and thorium in raw water by contacting rocks or soil containing natural radionuclides. Therefore, there is a possibility that workers in water treatment facilities will be exposed due to the accumulation of natural radionuclides in the water treatment process. The goal of this study is to evaluate the external radiation dose according to the working type of workers in water treatment facilities. In order to achieve the above goal, the study was conducted by dividing it into 1) analysis of the exposure environment, 2) measurement of the external radiation dose rate 3) evaluation of the external radiation dose. In the stage of analyzing the exposure environment, major processes that are expected to occur significantly were derived. In the measurement stage of the external radiation dose rate, a map of the external radiation dose rate was prepared by measuring the spatial radiation dose rate in major processes. Through this, detailed measurement points were selected considering the movement of workers. In the external radiation dose evaluation stage, the external radiation dose was evaluated based on the previously derived external radiation dose rate and working time. As a result of measuring the external radiation dose rate at the detailed points of water treatment facilities A to C, it was 1.90×10^-1 to 3.75×10⁰ μSv h^-1, and the external radiation dose was analyzed as 3.27×10^-3 to 9.85×10^-2 mSv y^-1. The maximum external radiation dose appeared during the disinfection and cleaning of activated carbon at facility B, and it is judged that natural radionuclides were concentrated in activated carbon. It was found that the external radiation dose of workers in the water treatment facility was less than 1mSv y^-1, which is about 10% of the dose limit for the public. As a result of this study, it was found that the radiological effect of external radiation dose of domestic water treatment facility workers was insignificant. The results are expected to contribute as background data to present optimized safety management measures for domestic NORM industries in the future.

• Applied Chemistry for Engineering
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• v.35 no.2
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• pp.140-147
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• 2024

In order to increase the utilization of biomass, an electrochemical performance was considered after manufacturing a carbon anode material (SV-C) for a Setaria viridis-based lithium ion secondary battery through a heat treatment process. When the heat treatment temperature of the Setaria viridis is as low as 750 ℃, the capacitance (1003.3 mAh/g, at 0.1 C) is high due to the negative (-) charge of oxygen present on the surface attracting lithium, along with the low crystallinity and high specific surface area (126 m²/g), but the capacity retention rate is believed to be as low as 61.0% (at 500 cycles and 1 C). In addition, it was confirmed that when the heat treatment temperature increased to 1150 ℃, the carbon layer was condensed to be excellent in arrangement, and the structural defects were reduced, resulting in a significant reduction in the specific surface area (32 m²/g) of the pores. Furthermore, when the surface defects of the anode material are reduced and the crystallinity is increased, the capacity retention rate is as high as 89.7% (at 500 cycles and 1 C), but the degree of defects is small, the active point is reduced, and the specific capacity is considered to be very low at 471.7 mAh/g. In the scope of this study, it was found that in the case of the Setaria viridis-based carbon anode material manufactured according to the heat treatment temperature, the surface oxygen content and crystallinity have higher reliability on the electrochemical properties of the anode material than the specific surface area.

• The Korean Journal of Nuclear Medicine Technology
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• v.14 no.2
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• pp.41-44
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• 2010

Purpose: Along with recent advances in PET/CT instrumentation and imaging technology, the number of patients has also been steadily increasing. This resulted in the increased radiation exposure to radiation workers in PET/CT rooms. In this study, we installed a radiation shield and investigated whether it could reduce radiation exposure to the workers and thus enhance job satisfaction. Materials and Methods: A radiation shield is composed of 5 cm thick lead and has a structure in which a radiation worker sits and watches a patient through lead glass while injecting radiopharmaceutical to the patient. Quarterly absorbed dose of radiation workers was measured using thermoluminescence dosimeters (TLD) and the results were compared for six months each before and after installation of the radiation shield. Exposure dose was also measured using a pocket dosimeter placed at the same location in the front and the back of the radiation shield. In addition, frequency of use of the shield and job satisfaction of radiation workers were investigated using a survey. Results: Quarterly absorbed dose of radiation workers was 2.70 mSv on average before installation of new radiation shield, whereas that dropped to 2.13 mSv after installation of radiation shield, reducing radiation exposure dose by 21%. Exposure dose on the front side of the shield was 61.2 R, whereas that on the back side of shield was 2.8 R. According to the survey, 85% of workers used the shield and were satisfied with the outcome: each radiation worker made injections to patients average of 6.5 times/day and preferred sitting to standing while injecting radiopharmaceutical to patients. Conclusion: Use of radiation shield reduced the exposure dose of radiation workers, which is the ultimate goal of radiation protection to minimize radiation exposure and is an appropriate method for the improvement of hospital working environment. Furthermore, we found that use of radiation shield not only relieves physical and psychological burden of radiation workers but also enhances job satisfaction. This result indicates that use of radiation shield is important for improvement of the radiation workers' job environment in terms of radiation protection.

• The Journal of Korean Society for Radiation Therapy
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• v.24 no.2
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• pp.107-114
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• 2012

Purpose: Unlike the existing linear accelerator with photon, proton therapy produces a number of second radiation due to the kinds of nuclide including neutron that is produced from the interaction with matter, and more attention must be paid on the exposure level of radiation workers for this reason. Therefore, thermoluminescence dosimeter (TLD) that is being widely used to measure radiation was utilized to analyze the exposure level of the radiation workers and propose a basic data about the radiation exposure level during the proton therapy. Materials and Methods: The subjects were radiation workers who worked at the proton therapy center of National Cancer Center and TLD Badge was used to compare the measured data of exposure level. In order to check the dispersion of exposure dose on body parts from the second radiation coming out surrounding the beam line of proton, TLD (width and length: 3 mm each) was attached to on the body spots (lateral canthi, neck, nipples, umbilicus, back, wrists) and retained them for 8 working hours, and the average data was obtained after measuring them for 80 hours. Moreover, in order to look into the dispersion of spatial exposure in the treatment room, TLD was attached on the snout, PPS (Patient Positioning System), Pendant, block closet, DIPS (Digital Image Positioning System), Console, doors and measured its exposure dose level during the working hours per day. Results: As a result of measuring exposure level of TLD Badge of radiation workers, quarterly average was 0.174 mSv, yearly average was 0.543 mSv, and after measuring the exposure level of body spots, it showed that the highest exposed body spot was neck and the lowest exposed body spot was back (the middle point of a line connecting both scapula superior angles). Investigation into the spatial exposure according to the workers' movement revealed that the exposure level was highest near the snout and as the distance becomes distant, it went lower. Conclusion: Even a small amount of exposure will eventually increase cumulative dose and exposure dose on a specific body part can bring health risks if one works in a same location for a long period. Therefore, radiation workers must thoroughly manage exposure dose and try their best to minimize it according to ALARA (As Low As Reasonably Achievable) as the International Commission on Radiological Protection (ICRP) recommends.

• Progress in Medical Physics
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• v.17 no.2
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• pp.89-95
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• 2006

The purpose of this study was to evaluate the radiation doses during CT transmission scan by changing tube voltage and tube current, and to estimate the radiation dose during our clinical whole body $^{137}Cs$ transmission scan and high quality CT scan. Radiation doses were evaluated for Philips GEMINI 16 slices PET/CT system. Radiation dose was measured with standard CTDI head and body phantoms in a variety of CT tube voltage and tube current. A pencil ionization chamber with an active length of 100 mm and electrometer were used for radiation dose measurement. The measurement is carried out at the free-in-air, at the center, and at the periphery. The averaged absorbed dose was calculated by the weighted CTDI ($CTDI_w=1/3CTDI_{100,c}+2/3CTDI_{100,p}$) and then equivalent dose were calculated with $CTDI_w$. Specific organ dose was measured with our clinical whole body $^{137}Cs$ transmission scan and high quality CT scan using Alderson phantom and TLDs. The TLDs used for measurements were selected for an accuracy of ${\pm}5%$ and calibrated in 10 MeV X-ray radiation field. The organ or tissue was selected by the recommendations of ICRP 60. The radiation dose during CT scan is affected by the tube voltage and the tube current. The effective dose for $^{137}Cs$ transmission scan and high qualify CT scan are 0.14 mSv and 29.49 mSv, respectively. Radiation dose during transmission scan in the PET/CT system can measure using CTDI phantom with ionization chamber and anthropomorphic phantom with TLDs. further study need to be peformed to find optimal PET/CT acquisition protocols for reducing the patient exposure with same image qualify.

• The Korean Journal of Nuclear Medicine Technology
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• v.12 no.3
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• pp.157-170
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• 2008

Purpose: This prospective study was conducted to reveal the haematological index change by low level radiation exposure in radiological environment our hospital workers. Materials and Method: We gathered the cumulative dose by Thermoluminenscent Dosimeters (TLD) over 9-yr period and examined hematological index counts change (RBC, Hb, Platelet, WBC, Monocyte, Lymphocyte, Neutrophilic, Basophilic, Eosinophilic) both occupational workers and controls. Of a total 370 occupational workers and 335 controls were compared. Results: This analysis has led to the following general observations 1) The average cumulative dose in male and female were $9.65{\pm}15.2\;mSv$, $4.82{\pm}5.55\;mSv$ respectively. 2) In both male and female, there were very low relationship between occupation period and cumulative dose (r< ${\pm}0.25$). 3) Occupation period was more increased, in male, WBC counts decreased and increased workers, RBC counts decreased workers were more than controls group (p<0.05). In female, WBC counts decreased and increased workers and W-eosino counts decreased workers were more than controls group (p<0.01). 4) Cumulative dose was more increased, in male, W-Lympho counts decreased workers and Platelet counts deceased workers were more than controls group (p<0.05). In female, W-lympho counts decreased workers and RBC counts decreased workers were more than controls group (p<0.05). Conclusions: We can find some kinds of blood index abnormal distribution in occupational radiation workers by comparing with controls. Occupational workers cannot avoid radiation exposure, in spite of the control it. Actually low level radiation adverse effect occurred not dose but probability. So workers must always try to reduce exposure by ourselves, furthermore as long as possible the government should provide rapidly that national system on radiation control for worker's health.

• Journal of the Korean Society of Radiology
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• v.15 no.6
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• pp.803-811
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• 2021

In this study, the exposure amount of IASCC test worker was evaluated by applying the process simulation technology. Using DELMIA Version 5, a commercial process simulation code, IASCC test facility, hot cells, and workers were prepared, and IASCC test activities were implemented, and the cumulative exposure of workers passing through the dose-distributed space could be evaluated through user coding. In order to simulate behavior of workers, human manikins with a degree of freedom of 200 or more imitating the human musculoskeletal system were applied. In order to calculate the worker's exposure, the coordinates, start time, and retention period for each posture were extracted by accessing the sub-information of the human manikin task, and the cumulative exposure was calculated by multiplying the spatial dose value by the posture retention time. The spatial dose for the exposure evaluation was calculated using MCNP6 Version 1.0, and the calculated spatial dose was embedded into the process simulation domain. As a result of comparing and analyzing the results of exposure evaluation by process simulation and typical exposure evaluation, the annual exposure to daily test work in the regular entrance was predicted at similar levels, 0.388 mSv/year and 1.334 mSv/year, respectively. Exposure assessment was also performed on special tasks performed in areas with high spatial doses, and tasks with high exposure could be easily identified, and work improvement plans could be derived intuitively through human manikin posture and spatial dose visualization of the tasks.

• Journal of Radiation Protection and Research
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• v.14 no.2
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• pp.10-17
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• 1989

A study for the assessment of dose given by outdoor radon to respiratory system has been carried out by making use of radon-cups containing CR-39 plastic track detectors. Detection efficiencies were determined by irradiation of the radon-cups in a standard radon chamber of known concentration. Thus determined detection factors of CR-39 plastic track detector in bare, open cup and filtered cup geometry are found to be 0.273, 0.0813 and 0.0371 $trmm^{-2}$/(37$Bqm^{-3}{\cdot}d$), respectively, which are chemically etched in 30% NaOH solution of $70^{\circ}C$ for 220 minutes. The outdoor radon concentrations measured at Taejeon(Chungnam National University) from May 1988 to March 1989 are in the range of 27.4 - 135.8 Bq/$m^3$(0.74 - 3.67pCi/l)by open cup and 16.7 - 143.9 Bq/$m^3$(0.45 - 3.89 pCi/l) by filtered cup, which yield overall annual average value of outdoor radon concentration of $70.8Bq/m^3$(1.91 pCi/l). Corresponding effective dose equivalent rate to respiratory system of ICRP standard man is assessed to be 520 nSv /h.

• Proceedings of the Korea Air Pollution Research Association Conference
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• 2003.05b
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• pp.67-68
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• 2003

라돈은 일반적으로 가장 잘 알려진 천연 방사성핵종 중 하나로서 무향 무색의 불활성기체이며 붕괴과정에서 알파입자를 방출한다. 라돈에 의한 피폭선량은 라돈붕괴에 의해 생성된 라돈자손이 호흡기관 표면에 침착되어 방출하는 알파선에 기인한다. 따라서, 피폭선량에 주로 기인하는 것은 라돈 자신이 아니라 그의 단 반감기 라돈자손들이다. 이처럼 라돈은 잘 알려진 폐암 유발원으로서 고농도의 라돈에 장기간 노출되는 경우 폐암을 유발할 수 있다. UNSCEAR 보고서(1993)는 자연 환경중에서 인간이 받는 연간 총 피폭선량인 2.4 mSv중 약 50%에 해당하는 1.15 mSv가 라돈과 그 자손에 의한 것이며 대부분 옥내에서의 호흡에 의해 비롯된다고 평가하고 있다. (중략)